foia/pa no: 2014-0024 records being released in their …some time later. additional inspection to...

FOIA/PA NO: 2014-0024

RECORDS BEING RELEASED IN THEIR ENTIRETY

Allen, DonA "e, o "Y1 "'~

From:Sent:To:Cc:Subject:Attachments:

Miller, GeoffreyThursday, May 09, 2013 3:45 PMUselding, LaraAllen, Don; Dricks, Victor; Kennedy, Kriss; Scott, Michael; Blount, Tom; Azua, RayANO AIT Public Meeting NotesANO AIT Script for Miller.docx; ATl"presANO.ppt

Lara,

Victor asked me to forward you the attached speaking notes from the meeting today. I may not have usedthese exact words, but I stuck pretty close to the script. I also attached the final version of the slides, though Idon't think they changed from the last one I sent out.

There were about 150 people present, including the family of the deceased (and their lawyer). Most of the restwere plant employees, with a couple newspapers and a TV camera (Victor gave an interview with them). Thepresentation lasted about 25 minutes, and there were only a few questions afterwards. Overall, I wassurprised at how short it was - I'm hoping that was because we were effective at getting our message out(guess we'll know when the newspapers come out).

Please let me know if you have questions. I'll be back in the office on Monday.

Thank you,

Geoff

I

ANO AIT Script for MillerSUMMARY

A. Opening Remarks1. Why we're here2. Why an AIT warranted3. What is an AIT?4. Logistics of the CAT-1 Meeting (agenda, feedback forms, public participation)

B. Summary of Inspection Results

1. SYNOPSIS OF EVENT

2. AIT ACCOMPLISHED ITS PURPOSE

3. REACTOR PLANT SAFETY SYSTEMS RESPONDED AS DESIGNED TO THE LOSSOF OFFSITE POWER AND UNIT 2 REACTOR TRIP

4. LICENSEE TOOK APPROPRIATE ACTIONS TO RECOVER PLANT EQUIPMENT ONUNITS 1 AND 2 AND HAS INITIATED AN EXTENSIVE CAUSE EVALUATION EFFORT

5. NRC RESPONDED PROMPTLY AND CONTINUES TO INSPECT

6. SUMMARY OF INSPECTION AND TEN UNRESOLVED ITEMS

7. FOLLOW-UP INSPECTION TEAM WILL REVIEW THE SIGNIFICANCE OF THESEURIs AND DETERMINE ANY ENFORCEMENT ACTION WARRANTED

C. QuestionslRemarks from Arkansas Nuclear One

D. Concluding Remarks by Kennedy

B. Summary of Inspection ResultsDETAILS

OPENING REMARKS

GOOD AFTERNOON. MY NAME IS GEOFFREY MILLER. I'M WITH THE NUCLEAR REGULATORYCOMMISSION, AND I AM THE TEAM LEAD FOR THE RECENTLY COMPLETED AUGMENTED INSPECTION

AT ARKANSAS NUCLEAR ONE. I'D LIKE TO START BY OFFERING SINCERE CONDOLENCES TO THE

FAMILY AND FRIENDS OF THOSE WHO INJURED OR KILLED BY THE EVENT ON MARCH 31. WE

RECOGNIZE THAT THIS EVENT HAD A SIGNIFICANT EMOTIONAL IMPACT ON THE PLANT AND

SURROUNDING COMMUNITY, AND THAT THERE IS UNDERSTANDABLY A GREAT DEAL OF INTEREST IN

THE CAUSES THAT LED TO THE EVENT. THE CAUSES OF THE INDUSTRIAL ACCIDENT ARE THE SUBJECT

OF AN ONGOING INVESTIGATION BY THE OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION

(OSHA). OUR MEETING TODAY WILL NOT INCLUDE A DISCUSSION OF THE CAUSES. OUR INSPECTION

FOCUSED ON THE EFFECTS THE EVENT HAD ON THE NUCLEAR PLANTS AT THE STATION AND THE

STEPS TAKEN BY OPERATORS IN RESPONSE TO PROTECT THE PUBLIC HEALTH AND SAFETY.

WE'RE MEETING TODAY WITH ENTERGY OPERATIONS TO PROVIDE A STATUS REPORT OF OUR

ONGOING INSPECTION ACTIONS. FOR MEMBERS OF THE PUBLIC WHO ARE IN ATTENDANCEAT THIS MEETING, NRC STAFF WILL BE AVAILABLE TO ANSWER QUESTIONS ANDRECEIVE COMMENTS AFTER THE BUSINESS PORTION OF THE MEETING.

WITH ME HERE TODAY.... [INTRODUCE THOSE IN ATTENDANCE INCLUDE VICTOR]

NOW, MR. BROWNING, WOULD YOU LIKE TO INTRODUCE YOUR STAFF?

ONE OTHER ADMINISTRATIVE ITEM: THERE ARE FEEDBACK FORMS AVAILABLE AT THEBACK TABLE. IN OUR CONTINUING EFFORT TO PROVIDE MORE MEANINGFULMEETINGS WITH OUR STAKEHOLDERS, WE WOULD APPRECIATE YOU TAKING THETIME TO COMPLETE ONE OF THE FORMS AND RETURN IT TO US. WE WILL USE YOURFEEDBACK IN OUR CONTINUING PROCESS TO IMPROVE THE QUALITY OF OURINTERACTIONS WITH OUR STAKEHOLDERS.

[REVIEW AGENDA]

SUMMARY OF THE INSPECTION RESULTS

1. AIT ACCOMPLISHED ITS PURPOSE

AUGMENTED INSPECTION TEAMS ARE USED BY THE NRC TO REVIEW MORE SIGNIFICANT EVENTS OR

ISSUES AT NRC-LICENSED FACILITIES. AN AUGMENTED INSPECTION TEAM IS USED WHEN THE NRCWANTS TO PROMPTLY DIG DEEPLY INTO THE CIRCUMSTANCES SURROUNDING AN OPERATIONAL EVENT

TO MAKE SURE THAT ALL OF THE CIRCUMSTANCES THAT CONTRIBUTED TO THIS EVENT ARE WELL

UNDERSTOOD IN ORDER TO PREVENT A RECURRENCE.

SINCE THIS EVENT INVOLVED MULTIPLE SYSTEM FAILURES, AND BASED ON OUR ESTIMATE OF THE RISK

INCREASE TO THE PLANT CAUSED BY THE EVENT, REGION IV CONCLUDED THAT THE NRC RESPONSE

2


SHOULD BE AN AUGMENTED INSPECTION TEAM. THE PURPOSE OF TODAY'S MEETING WILL BE TO

PUBLICLY PRESENT THE ITEMS IDENTIFIED BY THE INSPECTION TEAM AS POTENTIAL ISSUES REQUIRING

ADDITIONAL FOLLOW UP INSPECTION.

The NRC assigns full-time inspectors, called "resident inspectors," to each operating reactorfacility (ID Fred, Abin, William). The resident inspectors conduct daily inspections at ANO andlive in the surrounding community. Should an event occur at the plant, the resident inspectorsprovide immediate response capability for the NRC to assess plant conditions and licenseeactions. For this particular event, within one hour of the crane collapse, Fred and Abin were onsite monitoring operator actions and the safety of the reactors.

As I mentioned earlier, the purpose of an augmented inspection for NRC to promptly assessmore significant events and their causes; to gather the facts and identify issues that may beeither performance deficiencies or generic safety issues for the industry. This event resulted inwidespread equipment damage, including a loss of offsite power to a unit in a refueling outageand a trip and emergency declaration on the operating unit. Considering the equipment impactsand associated risk to the nuclear plants, an Augmented Inspection Team response wasappropriate.

The five-person inspection team consisted of experts in electrical, fire protection and operations,and a risk expert, with decades of experience in their disciplines.

The team spent more than a week on site with additional in-office inspection, conductedinterviews and physical inspections in the field, and reviewed system data and event records toindependently identify and understand all the issues that would warrant follow-up inspection.

THIS EVENT IS ALSO THE SUBJECT OF AN ONGOING INVESTIGATION BY THE OCCUPATIONAL SAFETY

AND HEALTH ADMINISTRATION. BOTH NRC AND OSHA HAVE JURISDICTION OVER OCCUPATIONAL

SAFETY AND HEALTH AT NRC-LICENSED FACILITIES. NRC AND OSHA HAVE A MEMORANDUM OF

UNDERSTANDING IN PLACE TO ENSURE A COORDINATED AGENCY EFFORT IN THE PROTECTION OF

WORKERS AND TO AVOID DUPLICATION OF EFFORT. THE OSHA INVESTIGATION IS STILL ONGOING,

WITH THE PRIMARY FOCUS BEING THE SAFETY AND HEALTH OF THE EMPLOYEES AND EMPLOYERS AT

THE FACILITY. THE NRC INSPECTION.THAT IS THE SUBJECT OF TODAY'S MEETING FOCUSED ON THE

IMPACT OF THE MARCH 31 EVENT ON THE EQUIPMENT AND SAFETY SYSTEMS ASSOCIATED WITH THE

TWO NUCLEAR REACTORS AT THE SITE TO ENSURE THE HEALTH AND SAFETY OF THE PUBLIC AND THE

ENVIRONMENT REMAINED PROTECTED FROM RADIOLOGICAL HAZARDS.

2. SYNOPSIS OF EVENT

THE EVENT THAT WAS THE SUBJECT OF THIS AUGMENTED INSPECTION OCCURRED ON MARCH 31WHEN A TEMPORARY LIFTING RIG BEING USED TO MOVE THE GENERATOR STATOR FROM UNIT 1

COLLAPSED, KILLING ONE PERSON AND INJURING EIGHT OTHERS. UNIT 1 WAS IN A REFUELING

OUTAGE AT THE TIME AND LOST ELECTRICAL POWER FROM OFFSITE DUE TO DAMAGE CAUSED BY THE

3


DROPPED STATOR. UNIT 2 WAS OPERATING AT FULL POWER AND AUTOMATICALLY SHUTDOWN WHEN

THE IMPACT OF THE STATOR ON THE TURBINE DECK CAUSED ELECTRICAL BREAKERS TO OPEN,

REMOVING POWER FROM ONE OF FOUR OPERATING REACTOR COOLANT PUMPS. WATER FROM A

RUPTURED FIRE MAIN LATER CAUSED A SHORT CIRCUIT AND SMALL EXPLOSION INSIDE AN ELECTRICAL

BREAKER ON UNIT 2, AND OPERATORS SUBSEQUENTLY DECLARED A NOTICE OF UNUSUAL EVENT

(LOWEST OF FOUR EMERGENCY CLASSIFICATIONS), TERMINATING IT AFTER TAKING CORRECTIVE

ACTIONS TO STABILIZE THE PLANT'S POWER SUPPLIES

Before we get into the specific details of the issues the team identified, I'd like to make a couple

general observations.

3. REACTOR PLANT SAFETY SYSTEMS RESPONDED AS DESIGNED TO THE EVENT

The team determined that after the event occurred, the plant safety systems responded asdesigned, that all assumptions in the accident analysis appropriately bounded the event, and no

unanalyzed condition existed. As such, there was no danger to the public health and safety

from radiological hazards.

4. ENTERGY HAS TAKEN APPROPRIATE ACTIONS TO RESTORE PLANTEQUIPMENT AND HAS INITIATED AN EXTENSIVE ROOT CAUSE EFFORT

To date, the Entergy response following the March 31 event appears appropriate. ANO

installed temporary modifications to restore offsite power to both units, and implemented

compensatory measures for security/fire protection; extensive RCE effort underway. They are

treating this event seriously as they determine causes and establish corrective actions. TheNRC will conduct additional inspection of the cause evaluation effort and the approach ANO will

use in prioritizing and implementing corrective actions. Lots completed, more work to come.

5. SUMMARY OF INSPECTION AND TEN UNRESOLVED ITEMS

The team was chartered by the Region IV Administrator to focus on several specific inspection

areas. I'll summarize the results of each inspection area:

1. Chronology of Significant Events.

We established a detailed Sequence of Events for the dropped stator event through therestoration of offsite power via temporary modifications. We did not identify any issues

requiring follow-up in this area

4


2. Operator Response.

Multiple challenges: personnel emergency, reactor trip, LOOP, fire water header break, lossof spent fuel pool cooling, breaker fault which led to the declaration of an UE. Operatorresponse appropriately protected the public health and safety.

URI #1: ANO's Control of a Modification Associated with Temporary Fire Pump- Temporary fire pump installed to augment the fire system during the outage.- Stator drop ruptured fire system piping in train bay and vicinity, causing significant

leakage into the train bay. DD pump started as designed to raise system pressure. Operatorsshut down the DD pump to stop the leakage, but did not shut down the temporary pump untilsome time later. Additional inspection to

3. Unit 1 and 2 Equipment Impact.

The team confirmed widespread damage to components within the turbine building[including fire barriers, fire doors, fire penetrations, fire piping, cardox piping, instrument airpiping, hydrogen piping, flood barriers, electrical cabinets and buswork, ventilation ducting,structural members.) Licensee assessment of damage is still in progress. A full assessmentwill not be possible until debris removal activities are completed. Additional follow upinspection as debris removal completed and areas become accessible. (URI #2: StructuralImpact to Units 1 and Unit 2)

4. Plant Response during the Event.

As I stated earlier, the team concluded the safety-related systems in Units 1 and 2responded as designed to the loss of offsite power and reactor trip, and that no unanalyzedconditions occurred as a result of this event. The team identified three items for furtherfollow up inspection:

URI #3: Control of Steam Generator Nozzle DamsThe nozzle dams are essentially inflatable plugs that are used to allow access to theinside of the steam generators for inspection during outages. At ANO, air pressure tomaintain the dams in place was provided by two separate electric air compressors.During the event, both air compressors lost power when offsite power was lost.Additional follow up inspection needed to review the methods used to provide airpressure to the nozzle dams.

URI #4: Main Feedwater Regulating Valve Maintenance Practices- MFRV stuck partially open during the last U2 scram due to a maintenance error. During

this event, the valve closed, but indicated open due to an indication problem from aseparate maintenance error, complicating operator response to the event. Additional

5


follow up inspection to review the valve maintenance. (ref NRC FIN 05000368/2012005,CR-2-2012-1432)

URI #5: Inadequate Flood BarriersAs discussed earlier, a considerable amount of water leaked into the train bay from a brokenfire main. The water leaked past flood barriers (gaskets in floor plugs) in the turbine buildingto the safety related auxiliary building and flowed to the aux building sump. Additionalinspection is needed to determine circumstances that allowed water to get from the turbinebuilding into the safety-related auxiliary building.

5. Compensatory Measures.

The team reviewed the adequacy of the licensee's compensatory measures for damagedequipment, including security barriers, support systems (equipment cooling) and fireprotection systems. The team concluded the licensee's compensatory measures wereappropriate and preserved plant safety. One item identified for further inspection associatedwith the timeliness of actions to restore water to the fire suppression system: (URI #5:Compensatory Measures for Fire Water System Rupture)

6. Event Classification and Reporting.

The team conducted an independent review of the licensee's actions for event classificationand reporting. The electrical fault on Unit 2 occurred at 9:23 in the morning, and an entry inthe station logs a short time later confirmed water intrusion and the failure of a breaker onthe associated electrical bus. Individuals from the field made several reports to the controlroom over the next hour (though none were logged), and operators declared a Notice ofUnusual Event at 10:33 a.m. The Emergency Action Level declaration was based on averbal report at approximately 10:20 a.m. of damage to the breaker consistent with a smallexplosion. The team concluded the identified Emergency Action Level (HU-4) wasappropriate. However, the team concluded additional inspection was required associatedwith whether the emergency declaration was timely based on the information available tothe control room. (URI #6: Timeliness of Emergency Action Level Determination)

7. Heavy Lift Preparations.

The team reviewed the licensee's plans and preparations for the movement of the stator,including their assessment of risk to the plant and identified an issue for further follow upinspection associated with the documentation of plant risk management administrative controlsfor the move. We identified a second issue for further follow up inspection associated with theevaluation of the vendor supplied crane per the licensee's material handling program. Thisissue will be examined as part of the licensee's root cause evaluation. NRC follow upinspection will be incorporated with the next charter item

6


8. Status of Cause Evaluation Efforts.

The team reviewed the licensee's initial efforts in establishing a cause evaluation team andthe beginning of the cause evaluation process. The root cause evaluation is still in progressat this time. We will conduct additional follow up inspection to assess the adequacy of thelicensee's identified causes and corrective actions when completed. (URI #9: Causes andCorrective Actions Associated with March 31, 2013, Dropped Heavy Load Event)

9. Operating Experience.

The team reviewed the licensee's application of operating experience, with specific focus oncontrol of heavy loads, contractor oversight, and seismic instrumentation. We expect plantsto review events from industry and incorporate lessons learned into their processes. Theteam concluded the licensee had appropriately incorporated the insights from industryoperating experience into their corporate programs and implementing procedures. The teamdid not identify any issues requiring follow-up in this area

6. FOLLOW-UP INSPECTION TEAM

That amounts to ten items requiring follow-up inspection that will be documented in this reportas Unresolved Items. The follow-up team will be assembled and dispatched after the details ofthe causes and corrective actions for these issues are identified. Their job will be to assess thesignificance of these issues and determine if any enforcement actions are appropriate.

[BROWNING][KENNEDY]Closed - Q&A

7

U.S.NRC

NRC Augmented Inspection TeamExit Meeting

Arkansas Nuclear One

Nuclear Regulatory Commission - Region IV

Russellville, AR

May 9, 2013

Agenda

" Introduction" Purpose of an AIT* Event Description" Inspection Results

• Licensee Response• NRC Closing Remarks

2 Protecting People and the Environment

Augmente Inspetio Tea

Fact-finding inspection

* Identify issues for follow up inspection

* Identify generic safety concerns in a timelymanner


March 31 Dropped Stator Event'*Ný

" Unit I RFO Stator Replacement" Structural Failure of Lifting Rig" Loss of Offsite Power to Unit I" Unit 2 Reactor Trip" Partial Power LosslBreaker Failure on Unit 2

" Notice of Unusual Event


Inspection Results

" Reactor plant safety systems responded asdesigned

* Entergy took appropriate actions to recoverplant equipment on Units I and 2 and hasinitiated an extensive root cause evaluationeffort

" NRC responded promptly and continues toinspect


Sumar ofCatetm

-,, ,,•.

" Event Chronology'

" Operator Response- Control of Temporary Modification

" Equipment Impact- Structural Impact to Units 1 and 2


Summary of Charter Items

* Plant Response to. the Event- Design Control of Steam Generator Nozzle

Dams- Main Feedwater Regulating Valve Maintenance- Flood Barrier Effectiveness

* Compensatory Measures- Fire Water Compensatory Actions

* Event Classification and Reporting- Timeliness of Emergency Declaration7 Protecting People and the Environment

0,10 Summary of Charter Items

Heavy Lift Preparations- Shutdown Reactor Equipment Risk-Implementation of Material Control Procedure

" Status of Cause Evaluation Efforts- Review of CauseslCorrective Actions

* Operating Experience

" Independent Risk Assessment Data8 Protecting People and the Environment

Licensee Response and Remarks

-----------


Contacting the NRC

" Report an emergency- (301) 816-5100 (call .collect)

* Report a safety concern- (800) 695-7403- [email protected]

• General information or questions- vvww.nrc.gov

- Select "What We Do" for Public Affairs


Electronic Distribution

• To receive a summary of this meeting andbeginreceiving other plant-specific e-mail distributions,subscribe to the Operating Reactor Correspondenceelectronic distribution via http://www.nrc.gov/public-involve/listserver/plants-by-region.html.

* To discontinue receiving electronic distribution, youmay unsubscribe at any time by visiting the sameweb address above.


Latta, Robert

From:Sent:To:Subject:Attachments:

MOSHER, NATALIE B <[email protected]>Thursday, August 22, 2013 9:58 AMWilloughby, Leonard; Melfi, Jim; Latta, RobertFW: ANO-1 LER 2013-001-01 Main Generator Stator Temporary Lift Assembly FailureOCAN081301.pdf

Thought you all would like a copy of the revised LER,

Sent: Thursday, August 22, 2013 8:42 AMSubject: ANO-1 LER 2013-001-01 Main Generator Stator Temporary Lift Assembly Failure

Outgoing NRC Correspondence

OCAN081301 -- dated 8/22/2013 - Subject: Licensee Event Report 50-313/2013-001-01 -ANO-1 Main Generator Stator Temporary Lift Assembly Failure

I

Latta, Robert

From: Willoughby, LeonardSent: Thursday, September 05, 2013 1:52 PMTo: Loveless, David; Allen, DonCc: Latta, RobertSubject: RE: ONE LAST TIME

I can live with the changes. Let's send it to HQ and tell them we their evalution done and back to us by COBOct 4, 2013.

From: Loveless, DavidSent: Thursday, September 05, 2013 10:31 AMTo: Willoughby, Leonard; Allen, DonCc: Latta, RobertSubject: ONE LAST TIME

I added a couple more comments I received. Please take one last look.

Thanks,

'- 'U.S.NRC

I~7JwtidgI•h •t £o¢ .'die~~w

Senior Reactor Ana.lt

(817) 200-1161

1

Latta, Robert

From: Willoughby, LeonardSent: Thursday, September 26, 2013 1:15 PMTo: Melfi, JimCc: Okonkwo, Nnaerika; Latta, RobertSubject: RE: AIT Followup Report

Jim,

At this time we are waiting on the SRA evaluation. Did you or Dave come to a conclusion on the recentlydiscovered geological features of the area that may affect flooding?

Please send me what you have on the SERP package.

Have fun in Chattanooga.

Leonard

From: Melfi, JimSent: Thursday, September 26, 2013 7:01 AMTo: Willoughby, LeonardCc: Okonkwo, Nnaerika; Latta, RobertSubject: AT Followup Report

Leonard

FYI, I will be in training the next 2 weeks in Chattanooga. (Oh Joy!!)

I will still be monitoring email, etc., on the ANO Followup report, and will review it when issued.

I have started SERP packages for the issues, but have not gotten very far.

What do you need from me for the report, or what can I assist you with from the region?

JIM

1

Werner, Greg -.... a

From;Sent:To:Cc:Subject:Attachments:

Weerakkody, SunilTuesday, November 19, 2013 11:47 AMLoveless, DavidWerner, GregANO LOOP Cut Set Report - M6-LOOP2 (ET) 2013_11_15 (2).rtfANO LOOP Cut Set Report - M6-LOOP2 (ET) 2013_11_15 (2).rtf

1

," % UNITED STATESNUCLEAR REGULATORY COMMISSION

WASHING'TON D 05A~-5-0001

ulfF !CEF OF T"E

-EA i.L CCONSEL

April 30, 2013

MEMORANDUM FOR. Reginald W. MitchellAssistant for Operations, OEDO

FROM: Andrew P. Averoach, SolicitorOffice of the General Counse~

RE: "LITIGATION HOLD' ON ALL MATERIALS PERTAINING TOACCIDENT RESULTING IN DEATH AT ARKANSAS NUCLEAR ONE

The Office of the General Counsel has been advised that the estate and family of Wade Waltersare likely to commence litigation related to the death of Wade Walters following a crane accidentat the Arkansas Nuclear One plant on March 31, 2013. We are also advised that a request hasbeen filed pursuant to the Freedom of Information Act for information pertaining to the NRC'sinvestigation of this accident, including any investigation reports, photographs, and inspectionreports. Therefore, it is appropriate to implement a 'Litigation Hold" on any documents relatedto the accident.

The implementation of a Litigation Hold requires NRC employees to.

1 Preserve any records related to the accident at Arkansas Nuclear One onMarch 31, 2013, including any documents generated as oart of theinvestigation of that accident. and

2. E-mail the name and contact information of any staff member likely to havediscoverable information to Andrew P. Averbach of OGC.

Preservation Duties:

1 NRC employeep .hniId takr mae.ureq teo preserve any materials relating to the

subject matter of the contemplated litigation. This obligation includes preserving "electronicallystored information" or 'ES! " NRC employees must preserve any electronically stored or writtenmaterial, wtrether final or in draft form, such as memoranda, e-mails, photographs, maps.diagrams, handwritten notes, databases, letters, presentation materials, recordings, microfilm,scanned photographs or documents. Working files may be kept in place, but must be identifiedon 'he inventory and readily retrievable

2: NRC employees may not delete, destroy, overwrite or mrow away potentiallyrelevant materials, including any relevant information in personal files, home computers orpersonal e-mail accounts. Even privileged materials must be preserved because a court mayneed to review documents to evaluate claims of privilege,

3. Any office identifying its possession of records subject to this Litigation Holdsl o old designate a contact person to facilitate coordinaeon. A list of records Inventoried should

-2-

be maintained and updated by this contact person. Each office should review inventoriedrecords for any claims of privilege or for the presence of Safeguards information, proprietaryinformation, or classified information.

Staff Likely to Have Materials:

OGC is also requesting that the name and contact information of any staff member likely tohave information reWevant to the accident or any associated inspection or investigation beemailed to Andrew P. Averbach of OGC.

Please direct any questions - and provide all information - to Andrew Averbach. You mayreach Mr. Averbach at 301-415-1956 or andrew [email protected].

Miller., Cwaff rey I

Fro~m. S ott Michavl 'sent-,Thursday. May D9, 2013 4:06 PM

-Huyckc, Doug;, Rosaes- Coopiar, cindCcc. Kennedy, Kri~ss Allen~, D~on; IB!cun Tomi; Sanchez~. Alf redt, Azua, Ray; i4~wolf, Amt Lmis

Roefet~ Nay, Midiaet, Mt4f jirn, J Fail anks~,Abin; Sthaijp, Wil liprT, L~d Eriie Utile,

Da~rrell; Reyn~lds,, Steven;, Cl~prkJeff V~el.Arion; P~ru~ett, Troy, Lund, Lotliw, Evans,Ntkhele,, Markfeyý' Michaeý Kalypr mn KV ~Weil, Jenny, Howell. Linda; Miller, 15-1offrey

Subje~RL ANO Week~ly StatUS Reportfor week ending May 10, 2M13Atameit ~ANO Update Week Ending, May 10 2GI 3,Rev~ 1.do(4

ReVISI~i to add key massoges from the Al exit.,,

Sm~t. Thursdiay, May 014 2013 21?P

To, Hiyck,~ Doug; Ro68Ies-C~oulr, CindyM" a iwe,~LwsRtr;HaM~~el M~lCVi KenMedy, KrL~$; Allen'Och; gkbunt, Tom;Sace rdHyfioI;Mi,

Subject: ANO Weekly 5ttr Repoq or Yekending May 10,, 2013

Subject reptit attached. P'er QEDO. this report is to be providod every ot~her week, so niext update wilf ýefo

,('87) 0G1462

OrF~IW1L USE ON~LY - SENW$~IVE INT R'NIA'L INFORMATION

ArkpnIsig Nucla OeD"o~ Stator Eve It

At 7ý50, a~m. (CýDT) an March, 31, 2013,, while lifting and trartsferrliN the Arkansas Nuclear OneUnit I main ganerator stator, t6 the-tralp 'ýay, theo lift systpm, colapsed, oaswing ttwe 525-ton-!"~or to fall w~an Ad'oerndey dmage potiohm of the tvirblnedeok and stibsequantly to! WIloveý 3 feet Into, it train, bray, At th tieo he even t. iit It was in aý refuelingi out age I Thereacor vessel head had been rer ovied, fuel ~was in the r~e~to ve~l and the ref, ue4ing,0aVityims flo0~ied up with walki.v"l greater thn23 feet above the fuel. Unit 2 was operatig~al

lbe' falure' of the liftirng dev'ide an~d the dropped stfnto clpmoed ý110resulting in a loss 6f oftte piowor to Unit 1.-Eerec i'ia geeaosstrted arnd mtorepower lolhe vital b~isses.i On Unit '2',the event caused' a reacotr motant pump breaker to opinmeulliing in a Unit,2 reactor trip from"1I 1o% power. L;Ser, du to fire water irvtrusion into Uni~t 2

swfthgear (ffie fire main'wia5 idamaged du~rvg tfte ovent) a'fftite Powe r Was lostl to ne, f the~Urtf2 vital busses duo toite ta~iure of a brealksTer assmiaad~ -Mrgec diesel generatorowlsed and reetored Po~r to the bus. The l~ceens." deeIared a Notification of Unusu~al Eventdue to the failure ýexpias~on) of the breaker..

Unit ICiwraril ftatus

:Sta tor av ald .rs~ h§* etmdwtvd Rcrvmipti &

'.. f1ite power ig, ovailablok via a torNprary modlifcation from, Startup Transformer I to vital

Power to nion-vital 480V buse is bQeflg supp!We by- ooliain fo. ~ oro ore

.Fire imawir ia pren'urizedi withr damag~ed sctionr.isolatd iswthsw np s

Unit 2 Curren~t St~tts

* Al~riate AC Diesel Goer0ator (Blackout Diesel,) suppily b ius h~s beeni repairttd ond can InodI

F Emergertcy dIesel qar~erators mf In standby,*Fire maln is pms~surized With darnagml sections isolat06.,Fifle Wathes ar ir.....* s

OYFFICIAL USE ONLY - SENSITIVE INTERNAL INFORMATiON

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OFFIC4AL USE ON .LY - $NSITE INTERN4AL1 .FORMATION.

Retorned A ffstew power to the yital busses on Aprl Ovia a, tempvtrly hQifiCation ftM

*Licensee's root cause teamn is on site and wo1rking on the evaluation* Th ~ tre up~~ ¶u e~pr~ ~f ~4e Lkt I { jrilsbIing~ The Ikenme

* Two QoftO~ poweor sou~rces andi the ;ýftetniae AC ieswei genorator hav4 b~er restored.,PotTp*asetRpr and r~arlresoluffio of items da'aqd itheev'ent were

rev:-wd th o ft Saet'keewC mminit! Apri"119.Th,* liimers anid tha NRC Wricpated in a Unit 2 phofteiall on April 24 to discoss tle,

ornsee~se rstart plan and related concemsI. i ..

'd Thsln ý pesidpnri ~ d- evaluat D

adtho ohii-ik~giey w ai &

SThe-resident inspectors corntiniie to n onitor,fiesee actions.

~~~portsgred Ah aLb~ At 'xtt e~n ku oPdnas~~y o ~ eet'so~~~JW~awm that~r the~a - b ltt ~i~owee etook

icbo~j~ ec~v ~et~uim~rt ~d h~uro oi446 01rie *ýW*n h"~ not

Res~dents, Regwion an NRR rwvie~egý the licenuel's60 evalullol to 04r¶y Viat the,tempom'ry -Dffsite power ýwmr satisfies the Vnft 1TehrWcal 5ecifi( atiori rmquiremntspriof to deWuei~rvg the reactor,

NRC continimueto respond to questions fromn the medila. -A press relese was Is-sue to'announo6 the *inning of the augmented insipec~tion and another to'annourice the public,extmeIyg

NROIO$HA 6oordlIiatIon

OFFICIAL USE ONLY -SENSITIVE.17NTERNAL NOM IN

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OPFtIAL USE ONLY - SENSITIVE INTERNAL INFORMATION:

NRCstaff and, OSHA stf otnet odnt cliuadsaeIfg Ftinspct

Manal haper 007andtheNRC10WMemoiaridurnf Uhderstandiing dated October

,i "•!,:•:!:•:.•!'.:•.: •:;:: 2 1:.. ,•:!": I w o,: :':: • •:•: " ::::> ::::. . :•.: ',:.. • •! • !i 7 i•-T i•:• 7 :F -•:i. F!:: % =•~ •-::'7ii::JP.i'::ii'••i"•~i :i,:

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• ~ ~~. . . . ... .. .. . . . . .. ... . ..

I OFFICIAL USE 5ONLY - SENSITIVE INThRNAL INFORMAMtN

i" " " . . . . . . ' :. . :. i .. . . . . .:... .:.Z . . " . . . . . . .. . .: . .. :. .. .. . . : .... • .. . . . .. . ... . . . . .. . .. . .. : :. . 7 . : : . i . ' ? . . .. :3

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Miller. Geoffrey ,,

From: Jones, SteveSent: Monday, April 29. 2013 2:51 PMTo: Miller, GeoffreySubject: RE: Preservation of ]nformation Related to Death of Wade Waiters at ANO

Geoff.

I recommend using "temporary overhead crane' because the definition of "overhead crane" in ASME B30 2 is:

A crane with a multiple-girder movable bridge carrying a movable or fixed hoisting mechanism and traveling onan overhead fixed runway structure.

In this case, the moveable bridge was the trolley on which the stand-lack hoists were mounted as a fixedhoisting mechanism. The overhead runway structure, although temporary, was fixed by attachments to theturbine building structure. A gantry crane is similar to an overhead crane, except the bridge is rigidly supportedon legs running on fixed rails. The ANO temporary crane had no legs between the girders and the wheeltrucks. The same standard applies to either design (ASME 830.2), but the legs on a gantry crane allowgeneration of much larger moments at the ends of the oridge. Since ASME B30.2 is cited in OSHAregulations, it would be better to just use overhead crane.

I received the FOIA earlier today- I do have a three procedures and the outage schedule. which I was holdingin case there was a need to discuss my input with Region IV management. Is the report in concurrence now?

Steve

From: Miller, GeoffreySent: Monday, April 29, 2013 3:25 PM .1 • /.:.fi KTo: Jones, SteveSubject: FW: Preservation of Information Related to Death of Wade Walters at ANOImportance: High

Steve.

Please see the below.

Thank you for getting me your timely input. We are planning an exit meeting on May 9 (you need not travel outfor that).

OSHA is referring to the lift device as a "temporary overhead gantry crane." I know we had a discussion that itwas a crane, but not a gantry crane. Would it be incorrect to use this term? Would it be better to go moregeneric (e.g., 'temporary overhead crane')?

Thanks,

Geoff

From: Tannenbaum, Anita On Behalf Of Fuller, Karla " ,j -

Sent, Monday, April 29, 2013 2:20 PMTo: HoWell, Art; Lewis, Robert; Kennedy, Kriss; Scott, Michael; Blount, Tom; Clark, Jeff; Farnholtz, Thomas; Miller,Geoffrey; Haire, Mark; Drake, James; Watkins, John; Kellar, Ray; Gaddy, Vincent; Allen, Don; Azua, Ray; Bradley, Dan;Melfi, Jim; Sanchez, Alfred; Schaup, William; Fairbanks, Abin; Hatfield, Gloria; Jones, Steve; Ahern, Gregory; Alexander,

i

Ryan; Afferink, Beth; Andrews, Tom; Diederich, Karl; Livermore, Dan; Makris, Nestor; Rodriguez, JaimeCc: Quayle, Lisa; Fuller, Karla; Lusk, Rustin; Lackey, Dana; Harrison, Deborah; Karl, TracySubject: Preservation of Information Related to Death of Wade Walters at ANOImportance: High

On April 23, 2013, we received a letter from a law firm representing the estate and family of Wade Walters related tothe recent industrial accident at the Arkansas Nuclear One plant. The firm requested that we preserve all our findings,reports, evidence, data, videos, and all information about the event. We have also received a request pursuant to theFreedom of Information Act for such material, In the near future, we will be circulating instructions concerning thecollection and preservation of this material- In the meantime, however, please make sure you preserve all informationthat is relevant to this event.

If any NRC employees were omitted from the distribution/addressee list who should receive this message, please shareIt with them immediately.

if you have any questions, please contact me. Thank you for your assistance in this matter.

Karla

Karla Smith Fuller, Esq.Regional Counsel

U.S. Nuclear Regulatory CommissionRegion IV1600 East Lamar Boulevard 0.3016Arlington, TX 76011

817-200-1271 (work)Karla. Fuller(M nrc.gov

t- ~~~'1P•J5Ob4?~

2

Miller, Geoffrey

From:

Sent:To:Cc:Subject:

.one., SteveTuesday, May 07, 2013 10:17 AMMiller, GeoffreySanchez, AlfredRE: REVIEW: ANO AIT Key Messages

Geoff,

I agree the statements are accurate. On the second bullet, multiple members of the licensee's organizationstated what you have in brackets.

Steve

From: Miller, GeoffreySent: Tuesday, May 07, 2013 9:38 AMTo: Jones, Steve; Sanchez, AlfredSubject: REVIEW: ANO AlT Key MessagesImportance: High

Fred/Steve,

Below are some key messages I'd like to use for the AIT exit meeting. Could you please take a took and tellme if the statements are accurate?

Thank you!

Geoff

* Reactor Plant Safety Systems Responded as Designed [point: safety-related stuff worked]

* No Documented Ties between the Heavy Load Lift and Other Outage Activities in the Station OutageRisk Plan [point: rig failure not considered credible risk]

Entergy Took Appropriate Actions to Recover Plant Equipment on Units 1 and 2 and Has Initiated anExtensive Root Cause Evaluation Effort [point: we don't have a failure cause yet]

I

Miller, Geoffrey-

From:Sent:To:Subject:

Hi Geoff,

Pannier. StephenTuesday, July 16, 2013 3:14 PMMiller, GeoffreyRE: Potential Generic Comms from ANO A]T ,- ,"~

I am still thinKing about how to proceed. As you know... pushing IN's into the pipeline is a daunting task. I hadthought about bundling all three and putting this out as one IN. I am not sure licensees read inspectionreports, but I know that they at least acknowledge receipt of an IN.

Thanks for asking. I'll be in touch.

Steve

'I, From: Miller, GeoffreySent: Friday, July 12, 2013 4:14 PMTo: Pannier, StephenSubject RE: Potential Generic Comms from ANO AIT

I

Steve,

Have you received word on whether or not any of the below topics would be suitable subjects for a new orupdated generic comm.?

Thank you for your help,

Geoff

From: Miller, GeoffreySent: Wednesday, June 05, 2013 8:40 AM

To: Pannier, StephenSubject: Potential Generic Comms from ANO AIT

w

Steve,

Below are some topics flagged as potential subjects for generic communications during the ANO AIT. Couldyou please take a look and let me know if there would be benefit in pursuing any of these based on existingcomms/OpE?

Thank you very much for your help,

Geoff

. Regarding the impact of trolley on Unit 2 turbine deck. potential generic communication associatedwith assessing scope of area when evaluating load paths (i.e., impact of heavy loads outside ofintended load path).

- Potential generic communication associated with integration of major project schedules for outagerisk assessment (IF confirmed to be an issue by AIT follow up inspection).

- Potential generic comm for diverse methods of supply for nozzle dams (N2 bottles) ,

. .... .. ............... .. ..... . ... .... ...

2

Werner, Gre g

From:Sent:To:Cc:Subject:Attachments:

Weerakkody. SunilTuesday, November 19, 2013 11:46 AMLoveless, David; Miller, Geoffrey; Werner, GregMitman, Jeffreyprelimianry draft of ANO Stator Drop SDP Draft Revison 0ANOI LOOP SDP Analysis Rev 0.0.docx

From: Mitman, JeffreySent: Thursday, November 14, 2013 4:32 PMTo: Weerakkody, SunilSubject: ANO Stator Drop SDP Draft Revison 0

Sunil, attached is the subject analysis and a zip file containing the SPAR model for review and comment. Thezip file may be too big to email to Region IV. I'm still working on the SDP on the loss of SDC at midloop,

Jeff Mitman*1L_ j/

2

tut Set Report - M6-LOOP2 (ET)Nov 15. 2013

PWR D SPAR MODEL FOR ARKANSAS NUCLEAR ONE UNIT I

- Probf TotalCuSt_0a431# ]Frq.Cut Set DescriptionFrg Displaying 50 of 43171 Cut

10184 * 61.5e4 1 UQ ~~yn- 1.___1____ I___ Sets.

I 6.46E.5 j 42.7 M6-LOOP2: 19

1.OoEtV IE-MO-LOOP LOOP Event Occurs aunng Mode 63.01 E-2 1 EPS-OGN-FR-OG1 DIESEL GENERATOR I FAILS TO RUN

3.01E-2 1 EPS-DGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN

714F. EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

2 2.92E.5 19.3 M6-LOOP2: 19

1 MOE+0 IE-M6-LOOP LOOP Evert Occurs during Mode 6

4 09E-4j EPS-DGN-CF-DG12R CCF OF DIESEL GENERATORS DGI&DG2 TO RUN

7.14E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DiESEL IN 72 HOURS

3 6.21E-6 4.11 NISLOQP2 :19

1.OOE+O IE-M6.LOOP LOOP Event Otzurs during Mode 6

3.01E-2 EPS-DGN-FR-0G1 DIESEL GENERATOR I FAILS TO RUN)_.89E-3 EPSDGN.FS-DG2 DIESEL GENERATOR 2 FAILS TO START

7.14E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

4 8T21E-6 4.11 M6-LOO12: 19

1.0OE to IE-M6-LOOP LOOP Event Oucurs during Mode 6

3.0'E-2 EPS-OGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN2.89E.3 EPS-DGN-FS-DG1 DIESEL GENERATOR 1 FAILS TO START7.14E-2 EP$-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

5 5.14E-6 3.4 M6-LOOP2.19

1 00E+0 IE-M6-LOOP LOOP Event Occurs during Mode 6

2.39E-3 ACP-CRB-OO-1A308 4160V AC BREAKER 152-308 FAILS TO CLOSE

3-01E-7 _ EPS-DGN-FR-DG2 DIE SEL GENERATOR 2 FAILS TO RUN


8 5.14E-6 3.4 MG-LOOP2: 191.00E÷0 IE-M6-LOOP LOOP Event Occur, during Mode 62.39E-3 ACP-CRB-OO-IA408 41 WV AC BREAKER 152-408 FAILS TO CLOSE

3-01E-2 EPS-DGN-FR-DG1 DIESEL GENERATOR 1 FAILS TO RUN

7.14E-2 EPS-XHE.XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

7 2-91E•i 1.92 M6-LOOP2:19

SI.O1E+O IE-M8-LOOP LOOP Event Oexurs during Mode 5

3.01E-2 EPS-DGN-FR-DGI DIESEL GENERATOR I FAILS TO RUN

7 74E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS9.93E-1 SWS&-4C-RUNNING SWS MDP P4C IS RUNNING' 4B ALIGNED TO RED TRAIN

1.36E-3 SWS-MDP-FS-P4C SERVICE WATER MDP P4C FAILS TO START

8 2.58E-0 1.71 M5-LOOP2: 191 .0QE+0 IE-MB>-LOOP LOOP Event Occurs during Mode 83.61E-5 EPS-DGN-CF-DG12S CCF OF DIESEL GENERATORS DGI&0G2 TO START


9 2,.5E-6 1.42 M6-LOOP2: 19

1.00E+0 IE-M6-LOOP LOOP Event Occurs during Mome 6

3.01E-2 EPS-DGN-FR-DG1 DIESEL GENERATOR 1 FAILS TO RUN

- 7.14E-2 ' EPS-XHE.XL.NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

1.00E-3 EPS-XHE-XR-DG2 OP FAILS TO RESTORE DIESEL GENERATOR 2

10 2.15E..6 1.42 MS-LOOP2: 19

1.700E0 IE-M6-LOOP LOOP Event Occurs during Modte 63.OIE.2 EPS-DGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN


Model Version: 8.19Model Date: 04/30/2009

Page I Software Saphire 8,0.9

Cut Set Report - M6-LOOP2 (ET) PWR D SPAR MODEL FOR ARKANSAS NUCLEAR ONE UNIT 1Nov 16, 2013

1.OOE-3 EPS-XHE-XR-DG1 OP FAILS TO RESTORE DIESEL GENERATOR 1

11 2.07E-6 1.37 M6-LOOP2: 19.O-OE0 IE-M6-LOOP LOOP Event Occurs auring Mode 6

__ 3.01E-2 EPS-DGN-FR-OG1 DIESEL GENERATOR I FAILS TO RUN9.63E-4 EPS-MOV-CC-CV3&07 SYVS SUPPLY MOV CV-3W07 TO OGN 2 COOLING FAILS TO OPEN1.-14E.2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

12 2.0TE-6 1.37 M6-LOOP2 191-OOE±0 IE-W6-LOOP LOOP Event Occurs dluring Mode 63.01E-2 EPS-OGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN

g.63E-4 EPS-MOV-CC-CV3806 SWS SUPPLY MOV CV-3806 TO DON 1 COOLING FAILS TO OPEN7.14E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

13 t,69E-6 1,12 M6-LOP12: 19I OOE+O I IE-MS-LOOP LOOP Event Occurs during Mode 6

2.37E-5 EPS-MDP-CF-PI6ABS CCF of EDG Fuel Oil Pump to Start7.14E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

14 1.33E-6 088 M6-LOOP2: 191,O0E+0 : IE-M6-LOOP LOOP Event Occurs diing Mode 6

" 1.86E-5 EPS-MOV-CF-SWS CCF OF SWS SUPPLY MOVs 3806 AND 3807714E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

1 5 i.22E- o 08 M6-LOOP2LM4I .OOE.O IE-M6L LOOP Event Occurduring Mode 6

3.01E-2 EPS-OGN-FR-OGI DIESEL GENERATOR I FAILS TO RUN9.63E-4 I LPI-MOV-CC.CV1400 LPI DISCHARGE MOV CV-1400 FAILS TO OPEN

4.20E-2 _ LTREC-DHR-SD Late Recovery ol SDC/DHR (5 Days)

16 1.22E-6 0.8 M6-LOOP2: 041.O0E+0 IE-MB-LOOP LOOP Event Occurs during Mode 63.01E-2 EPS-DGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN9.63E-4 LPI-MOV-CC-CV1401 LPI DISCHARGE MOV CV-1401 FAILS TO OPEN

4 20E-2 LTREC-DHR-5D LW. e Recovery of SDCIDHR (5 Days).

17 1.20E e 0.70 M8-LOOP2; 04

1 .OOE+0 IE-MS-LOOP LOOP Event Occurs during Mode 6

3.01E-2 EPS-DGN-FR-DG1 DIESEL GENERATOR I FAILS TO RUN

4,20E-2 LTREC-DHR-SD Late Recovery of $DC/DHR ý5 Days)

9.51E-4 SWS-AOV-CC-CV'3841 FAILURE OF SWS MOV CV-3841 TO PMP P34A TO OPEN

16 l.20E-o 0.79 M6-LOOP2, 04 ......1, 00E0 0 IE-MB-LOOP LOOP Event Occurs during Mode 63.O1E-2 EPS-,IGN-FR-L)GZ ,)IESEL GENEHAIOR 2 FAIL, 10 HUN

4.20E-2 LTREC-DHR-50 Late Recovery of SDCQDHR (5 Days)9.51E-4 SWS-AOV-CC-CV384C FAILURE OF SWS AOV CV-3840 TO PMP P34A TO OPEN

19 1.20E-6 0,79 M6-LOOP2 ; 041,0OE+G IE-MS-LOOP LOOP Event Occurs during Mode 6

3.01E-2 . EPS-DGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN

9.47E-4 LPI-MDP-FS-P34A LPI MDP P34A FAILS TO START

4,20E-2 LTREC-DHR-6D Late Recovery of SDCIDHR (5 Days)

20 1.201E 0,79 M-LOOP2-041,00E+0 IE.MB-LOOP LOOP Event Occurs during Mode 6

3.01E-2 EPS-DGN-FR-DG1 DIESEL GENERATOR I FAILS TO RUN

9.47E-4 LPI-MDP-FS-P34B LPI MOP P34B FAILS TO START4.20E-2 1 LTREC-DHR-50 I Late Recovery of SDCJDHR (5 Days)

21~,104E-6 0.69 M6-LOOP2: 04 __

1.00E*0 ± IE-M6-LOOP i LOOP Event Occurs during Mode 6 '

Model Version. 8.19Model Date: 04)30/2009

Page 2 Software Saphlre 8.0.9

Clut Set Report - M6-LOOP2 (ET) PWR D SPAR MODEL FOR ARKANSAS NUCLEAR ONE UNIT INov 1•, 2013 __.

4.20E-2 LTREC-DHR-5D Late Recovery of SDC/DHR (5 Days)

2.48E-5/ - SWS-AOV-CF-CV38401 CCF OF SWS AOVs CV-3840,341 TO PUMPS P34A/B TO OPEN

2z 9.96E.? (Jti.b NdI6-LOOP2 .04 ____________________________________

l.OOE-0 " E-M-LOOP LOOP Event Occurs during Mo.. 62.37E-5. LPI-MOP-CF-STRT LPI PUMP COMMON CAUSE FAILURES TO START

4 20E-2 L7REC-DHR-50 Late Recovery of SDC/DHR (5 Days)

23 9.03E-7 0.6 M-LOOP2 19

1 .00E40 IE.MS-LOOP LOOP Event OCOurs dudng Mode'6

1.26E-5 EPS-MDP-CF-PI6ABR CCF ol EOG Fuel O Pump to Run

7. 14E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

24 7.70E-7 0.51 M6-LO0P2G04

.00E+0 IE.M6-LOOP LOOP Event Occurs during Mode 6

1 S93E-5 LPI-ACX-CF-VCIXR Common Cause lailure of DHR Unit Coolers VUC-1A,1B, IC & 10 to RUN

4.20E-2 LTREC-OHR-5D Late Recovery of SDCIDHR (5 Days)

25 5.97E-7 0.39 M6-LOOP2: 90 O 0E+0 IE-M6,-LOOP ... ,LOOP Event Occurs during Mode 6•

2 8ME-3 i EPS-DGN-FS-DGI DIESEL GENERATOR 1 FAILS TO START

2.89E-3 EPS-DGN.FS-DG2 DIESEL GENERATOR 2 FAILS TO START

7.14E-2 EP,-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

,26 531E-7 0,35 MS-LOOP2: 04

--.00E+0 IE-M6-LOOP LOOP Event Occurs during Mode 6

1 .26E-5 LP!-MDP-CF-RUN LPI PUMP COMMON CAUSE FAILURES TO RUN

4.7QE--2 LTREC-OHR-5D Late Recovery oi SDCJDNR (5 Days)

[27 4.94E-7 033 M6-LOOP2 :19I OC-E+O IE-M5,LOOP LOOP Event Occurs during Mode 6 "

2 39E-3 ACP-CRB-OO-1A308 4160V AC BREAKER 152-308 FAILS TO CLOSE ........... . ....

2.89E-3 EPS-DGN-FS-DG2 DIESEL GENERATOR 2 FAILS TO START


28 4.94E-7 0,33 MS-LOOP2: .19

1.0QE*-•Q IE-MO-LOOP LOOP Event Ociars duing MW!e 6

2.39E-3 ACP-CRB-OO-1A4 .. 4160V AC BREAKER 152-408 FAILS TO CLOSE

2.89E-3 EPS-DGN-FS-DG1 DIESEL GENERATOR 1 FAILS TO START

7.14E.& EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

29 4.58E-7 j 0.3 M6-LOOP2 : 04

1.00E+0 IE-MB-LOOP LOOP Event Occurs during Mode 6


3.62E.-4 i. LPI-MDP-FR-P34A LPI MDP P34A FAILS TO RUN

4.20E-2 LTREC-DHR-50 Late Recovery of SDC/OHR (5 Days)

30 4.58E-7 0.3 M6--LOOP2: 04

1.00E+0 J IE-MS-LOOP LOOP Event Occurs during Mode 6

3.01E-2 1 EPS-DGN-FR-DG1 DIESEL GENERATOR I FAILS TO RUN

3.62E-4 LPI-MDP-FR-P34B LPI MDP P348 FAILS TO RUN

4.20E-2 T LTREC-OHR-5D Late Recovery of SDC/DHR (5 Days)

31 4,09E-7 0.27 MB-LOOP2 19 . . ' ...............

1.00E+0 IE-M6-LOOP LOOP Event Occurs during Mode 6

2.39E-3 ACP-CRB-OO-1A308 4160V AC BREAKER 152-30e FAILS TO CLOSE

2.39E-3 I ACP-CRB-OO-1A408 4160V AC BREAKER 152400 FAILS TO CLOSE7-14 E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

32 2.79E-7 0.18 M6-LOOP2:19 _

_ 1.00E+0 IE-MW-LOOP LOOP Event Occurs during Mode6.

2,89E-3 . EPS-DGN-FS-DG1 D ESEL GENERATOR 1 FAILS TO START

Model Version: 8.19Model Date: 04130/2009


Cut Set Report - MG-LOOP2 (ET) PWR D SPAR MODEL FOR ARKANSAS NUCLEAR ONE UNIT INov 15 3 20t3

__ 7.14E2 _ EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS9.93E-1I SWS-4C.RUNNING 'M MDP"P4C IS RUNNING; 4B ALIGNED TO RED TRAIN

1.•1 E- i ,W--M0P.FS-P4C SERVICE WATER MOP P4C FAILS TO START

33 2.63E-7 0.17 M6-LOOP2 ý 19

t.00E+0 IE-M6-LOOP LOOP Event Oocum during Mode 6

7,14E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

9,93E-1 SWS.4A-RUNNING MWS MOP P4A IS RUNNING; 4B ALIGNED TO RED TRAIN

9.93E-1 SWS-4C-RUNNING SWS MOP P4C IS RUNNING: 48 ALIGNED TO RED TRAIN

3.73E-6 SWS-MOP-CF-STRT4A8C CCF OF SERVICE WATER MDPS 4A,48 & 4C TO START

34 2.31E-7 0.15 MG-LOOP2 .19

1.00E40 IE-MG-LOOP LOOP Event Occurs during Mode 612.30E-3 ACP-CRS-OO-IA308 4160V AC BREAKER 152-308 FAILS TO CLOSE7.i4E-2 EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

9.93E- " SWS-84C-RUNNING SWS MOP P4C IS RUNNING: 4B ALIGNED TO RED TRAIN

1.36E-3 SWS-MDP-FS-P4C SERVICE WATER MOP P4C FAILS TO START

35 2.06EJ7 0.14 MS-LOOP2: 19

1.E.0 IE-M6-LOOP LOOP Event Occurs dudng Mode 6

2 89E-3 EPS-DGN-FS-OG1 DIESEL GENERATOR 1 FAILS TO START

7 14E-2 EPS-XHE-XL.NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS' 1.00E-3 EPS-XHE-XR-DG2 OP FAILS TO RESTORE DIESEL GENERATOR 2

36., 2.06E-7 0 14 MS-LOOP2 19

1 OOE-0 IE-M6-LOOP LOOP Event Ocwurs during Mode 6

2.e9E-3 EPS.DGNFS-DG2 DIESEL GENERATOR 2 FAILS TO START


1 .OE-3 EPS-XHE-XR-OGI OP FAILS TO RESTORE DIESEL GENERATOR 137 1.99E-7 0.13 M6-LOOP2 : 19


2.89E-a EPS-DGN-FS-DG1 DIESEL GENERATOR I FAILS TO START

9.63E-4 EPS-MOV-CC-CV3807 SWS SUPPLY MOV CV-3807 TO DOGN 2 COOLING FAILS TO OPEN

3 T14E-2__ EPS-XHE-XL-NR7ZH OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 12 HOURS38 1 99E-7 0•,13 Me.LOOP2 :1

T1 ,00+0 T IE-MrLOOP LOOP Event Occurs during Mode 62.89E-3 EPS-DGN-FS-DG2 DIESEL GENERATOR 2 FAILS TO START

9.63E-4 EPS-MOV-CC-CV3806 SWS SUPPLY MOV CV-3806 TO DGN I COOLING FAlLS TO OPEN7.14E-2 _ _EPS-XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

39 1,90E-7 0,13 M6-LOOP2: 01.00E+0 __ IE-M&LOOP LOOP Event Occurs during Mode 6

3.33E-5 _ _ACP-BAC-LP-LCCB5 FAILURE OF LCC BUS ...

____ 3.(tE-2 _ _EPS.DGN.FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN

1.90E-1 LTREC-DHR-3D Late Recovery of S[DiT)HR 13 Days)

40 181E-7 .12 M6-LOOP2 ý 19

1 OOE,0 IE-M5-LOOP LOOP Event Occurs dunng Mode 6

3X1E-2 EPS-DGN-FR-DG1 DIESEL GENERATOR 1 FAILS TO RUN.. , ,.. . .. .. . . . . . . . . . .. . . . . . . . . . . . . ................... ... I.... 1.... ................ ......... I...... ...... . ....... ....................... ........ .. ...... .................... . ... . ....... ........... ..... ...

7.14E-2 EPS.XHE-XL-NR72H OPERATOR FAILS TO RECOVER EMERGENCY DIESEL IN 72 HOURS

9.93E-1 SWS-4C-RUNNING I SWS MDP P4C IS RUNNING, 46 ALIGNED TO RED TRAIN

8,47E-5 SWS-MOP-FR-P4C SERVICE WATER MOP P4C FAILS TO RUN41 1.71E-7 0.11 M6-LOOP2:19 _

IIE-_-LOOP_ LOOP Event Occurs durng Mode 6

2.39E-3t ACP-CR8-OO-1A308 4160V AC BREAKER 152.306 FAILS TO CLOSE7 14F-2 EPS-XHF,-XI..NR7?H OPERATOR FAILS TO RECOVER EMERGENCY DIESFI.. IN 72 HOURS

I ,0E3 ... EPS-XHE-XR-DG2 OP FAILS TO RESTORE DIESEL GENERATOR 2

Model Version, 8.19Model Date: 0413012009


Cut Set Report - M6-LOOP2 (ET) PWR D SPAR MODEL FOR ARKANSAS NUCLEAR ONE UNIT INov 15, 2013

42 1,71E-7 0.11 o A.LOOP2: 191.OOE+O IE-M6-LOOP LOOP Event Occurs during Mode 6

2.39E-3 ACP-CRB-OO-1A408 4 4160V AC BREAKER 152.408 FAILS TO CLOSE


~i 1.OCE-3 4 EPS-XHE-XR-DG1 OP FAILS TO RESTORE DIESEL GENERATOR 143 1 65E-7 0.11 M6-LOOP2 19

1.00E40 i IE-M6.LOOP LOOP Event Occurs during Mode 6

2.39E-3 ACP-CRB-OO-1A'08 4.60V AC BREAKER 152-308 FAJLS TO CLOSE

9.63E-4 EPS-MOV-CC-CV3807 SWS SUPPLY MOV CV-3007 TO OGN 2 COOLING FAILS TO OPEN


44 1.65E-7 0.11 MO-LOOP2. 19


2.39E-3 ACP-CRB-OO1,A408 41 60V AC BREAKER 152-408 FAILS TO CLOSE

9.63E-4 EPS-MOV-CC-CV3806 SWS SUPPLY MOV CV-3806 TO DGN I COOLING FAILS TO OPEN


45 1 .291-? 7 09 M&-IOP2 ' 081.0-0&0 IE-M6-LOOP LOOP Event Occurs during Mod.e .

2.27E-5 ACP-TFM-FC-X5 4160VI480V TRANSFORMER X5 FAILS


1.90E-1 LTREC-DHR-3D Late Recovery of SDC/DHR (3 Days)

46 1.29E-7 0.09 M6-LOOP2: 08

1OE+<0 IE-tM-LOOP LOOP Event Occurs during Mode 62,27E-5 ACP-TFM-FC-X6 4160V/040V TRANSFORMER X6 FAILS

3.011E-2 EPS-0GN-FR-OG1 DIESEL GENERATOR I FAILS TO RUN

1.90E-1 LTREC-DHR.3D Late Recovery of SDCJDHR (3 Days)

47 1.22E.7 0 08 WM6LOOP2 '_........... ............... . ...

I 00E+0 I IE-MW-LOOP LOOP Event Occurs during Mode 6

2.69E-6 LPI-ACX-CF-VCIXS Common Cause failure ot OHR Unit Cooiors VUC-lA, IB, IC & ID to Start

4.20E-2 LTREC-DHR-5D Late Recovery of SDC!DHR (5 Days)48 1.20E-7 008 !Mr.LOOP2 04 -

1.00E+0 _ IE-M&-LOOP LOOP Eent Occurs during Mode 6

- j 3 01 E-2 EPS-DGN-FR-DG2 DIESEL GENERATOR 2 FAILS TO RUN

9-50E-5 _ LPIACX-CF.VC1ABR Common Cause failure of OHR Unit Coolers VUC- 1A and 1B to Run

4.20E-2 LTREC-DHR-5D . Late Recovery of SDC/DHR (5 Days)

49 1. o0•E7 -F 0.08 M6-LOOP21 04 . ___.....___....___..........._lll___ lll___ .... ____ _

1.00E+0 I IE-M6-LOOP LOOP Event OCcurs during Mode 63.011E.2 EPS-DGN-FR-DG1 DIESEL GENERATOR 1 FAIL$ TO RUN

9.50E-.5 LPI-ACX-CF-VC1CDR Common Cause failure of DHR Unit Coolers VUC-1C and 10 to Run

4.20E-2 LTREC-0HR-50 Late Recovery of SDC(DHR (5 Days)

50 1 17E-7 j0.08 M6-LOOP2:04 __,

. 00E+0 IE-M-LOOP LOOP Event Occurs durin Mode62.89E-3 ; EPS-DGN-FS-G 1 DIESEL GENERATOR 1 FAILS TO START

9.63E-4 LPI-MOV-CC-CV1400 LPI DISCHARGE MOV CV-1400 FAILS TO OPEN4,20E-2 LTREC-DHR-5D Late Recovery of SOC/DHR (5 Days)

Model Version: 8,19Model Date: 04/3012009


(K~) 1)4~~~ AII { /Wemer, Greg

From:SentTo:

Cc:Subject:Attachments:

Weerakkody, SunilTuesday, November 19, 2013 11:46 AMLoveless, David; Miller, Geoffrey; Werner, GregMitman, Jeffreyprelimianry draft of ANO Stator Drop SDP Draft Revison 0ANO LOOP SDP Analysis Rev 0,0,docx

From: Mitman, JeffreySent: Thursday, November 14, 2013 4:32 PMTo: Weerakkody, SunilSubject: ANO Stator Drop SDP Draft Revison 0

Sunil, attached is the subject analysis and a zip file containing the SPAR model for review and comment. Thezip file may be too big to email to Region IV. I'm still working on the SDP on the loss of SDC at midloop.

Jeff Mitman

1

U.S.NRCUNITED STATES NUCLEAR RIEGULATORY COMMISSION

Protecting People and the Environment

Phase 3 Risk AssessmentLoss of Offsite Power

Arkansas Nuclear One Unit I

Revision 0.0

Probabilistic Risk Assessment (PRA) Analyst: Jeff Mitman, Senior Reliability and RiskAnalyst, NRR/DRA/APOB

Independent ReviewerRegion IV Reviewer

1.0 Introduction

On March 31 t 2013 at 7:50 am Ark~ansas Nuclear One Unit 1 (ANO1) experienced aloss of offsite power. While lifting and transferring the Unit 1 main generator stator to thetrain bay, the lift system failed, falling on to the turbine deck and into the train bay. Thisresulted in damage to the turbine building including damage to electrical buses supplyingoffsite power to Unit1 and damage to the fire suppression piping.

At the time of the event, Unit 1 was in a refueling outage. It had been shutdownapproximately 7 days. Fuel in the reactor vessel, the reactor cavity was flooded up, andboth trains of decay heat removal system were in service, With the loss of offsite power,both Unit 1emergency diesel generators started and loaded their respective buses.Decay heat removal (DHR) was quickly restored. Once DHR was restored the unit wasquasi stable with no offsite power available due to damage to the non-vital electricalbuses, with EDGs powering the vital busses and the decay heat removal systemoperating and providing decay heat removal to the reactor vessel.

Dropping the generator stator caused the following damage:

* Offsite power was lost - it took six days to recover* The station blackout diesel generator's (called the AAC) connection to the plant

was severed rendering the ACC non-functional% Fire watering piping was damaged requiring shutdown of the fire protection

system. It also caused flooding the Unit 1 and 2 structures with tens ofthousands of gallons of water challenging critical equipment

2.0 Discussion of the Performance Deficiency

The licensee failed to properly implement Engineering Procedure EN-MA-1 19, OMaterialHandling Program." The following two examples are presented:

SThe licensee failed to adeouately review and approve Bi ge Calculation 27619-Cl asrequired by Section 5.2f71(a)

Engineering Procedure EN-MA-1 19, Section 5.2[71 requires temporary hoistingassemblies to be designed or approved by Engineering Support Personnel (ESP). OnSeptember 12, 2012 Siemens Energy transmitted to Entergy, Bigge Calculation 27619-C1. "ANO Stator Replacement Project." The design calculation did not adequatelyconsider the loads that would be experienced by the lift. Entergy's review and approvalprocess failed to identify the calculation deficiencies and the weak component in thenorth tower structure. Specifically, Entergy's ESP failed to adequately review andidentify the flaw in Calculation 27619-Cl consistent with the requirements of procedureSection 5.2171(a) which states that temporary hoisting assemblies are required to bedesigned or approved by ESP. Had ESP's approval process identified the deficiencyand eliminated it prior to use of the assembly at ANO, the event would have beenavoided.

Page 2

* The licensee failed to ensure that a load test of the assembly to at least 125 percent ofthe projected hook load or to another approved standard was performed as required bySection 5.2[71(b) and associated note

Engineering Procedure EN-MA-i 19, Section 5,2[7](b) requires assembly's to be loadtested and held for at least five minutes at 125 percent of actual load rating before initialuse, However, the note in Section 5.2[71 allows specially designed devices, for specificapplications to be designed and tested to other approved standards. On February 14.2013 Entergy's supplemental Project Civil Engineer reviewed the letter from Bigge toSiemens Energy, dated February 8, 2013, and included that letter into EngineeringChange Notice ECN-39028. The Entergy engineer failed to identify that the uppercolumns and intermediate header were listed as new, negating Bigge's assertion that,"This hoist assembly has been used at other electric power stations to lift componentsthat exceed the anticipated weight of the unit 1 stator" This erroneous information wasthen used in lieu of a load test or testing in accordance with other approved standardsHad engineering personnel identified the erroneous information and a load test, ortesting to other approved standards, been performed, the deficiencies in the designwould have been detected prior to use at ANO and the event would have been avoided.

3.0 Plant Conditions Prior to the Event

Plant equipment and conditions were as follows:

" Unit was shutdown with fuel in the reactor, head removed and refueling canalflooded

" Estimated time to boil (TTB) was 8 hours" Estimated time to core uncovery was 3 days" Both trains of SDC were in service" Plant electrical lineup was in a plant shutdown configuration to support

maintenance and testing as follows:o 6900 Volt Bus H2 was de-energized.o 6900 Volt Bus Hi was energized.o 4160 Volt Bus A2 was de-energized.o Safety related 4160 Volt Buses A3 and A4 were cross tied and supplied

power via Non-safety related 4160 Volt bus Al.o 460 Volt buses Sb and l3ti were cross tiea.o Green Train battery D06 had been disconnected from D02 bus.o D04 battery charger was supplied from Swing MCC B56 to provide power

to Green Train DC bus D02.o B56 was aligned to B5.

Page 3

4.0 Plant Conditions after Initiating Event Initiated

Time to boil was estimated at eight hours and time to core uncovery without mitigationwas estimated at three days.

The following equipment was unavailable after event initiation:

* Offsite power" Station blackout diesel generator - ACC" Fire water" All balance of plant equipment" Gravity feed from the BWST as water level in the BWST was

lower than water level in RCS• Instrument air (IA) was unavailable - the analyst assumed that all

air operated valve failed in a safe direction, i.e.. the systems IAsupported remained available

• Starting air compressors for the emergency generators" Normal lighting

The following equipment was available after the event initiation to mitigatethe event:

* Both emergency diesel generators and their respective electricaldistribution systems

" Both decay heat removal trains (two pumps)" Both high pressure injection (HPI) trains (three pumps)" Reactor building spray systems - note these were not credited in

the analysis, however, the non-crediting had no effect on thequantitative results

5.0 Significance Determination Process (SDP) Phase 2 Summary

No Phase 2 was conducted.

6.0 Initiation of a Phase 3 SDP Risk Assessment

A Phase 3 SOP risk assessment was performed by the Office of Nuclear ReactorRegulation (NRR).

The analysts used the following generic references in preparing the risk assessment:* NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown

Management." December 19910 NUREG/CR-5883, "The SPAR-H Human Analysis MeLhod," August 20050 NUREG-1 842, "Good Practices for Implementing Human Reliability Analysis." April

2005• NUREG/CR-6595 Revision 1, "An Approach for Estimating the Frequencies of

Various Containment Failure Modes and Bypass Events." October 2004* INL/EXT-10-18533 Revision 2, 'SPAR-H Step-by-Step Guidance." May 20110 "RASP Manual Volume 1 - Internal Events," Revision 2.0 date January 2013* NUREG/CR-1278, "Handbook of HRA with Emphasis on Nuclear Power Plant

Applications," August 1983

Page 4

The analyst used the following plant specific references:

" EOP: 1202.007, Degraded Power* AOPs:

o 1203.024, Loss of Instrument Airo 1203.028, Loss of Decay Heat Removalc. 1203.050, Unit I Spent Fuel Pool Emergencies

" Calculation: 89-E-001 7-01, Time to Boiling and Time to Core Uncovery after Lossof Decay Heat Removal, Unit 1, Revision 7

" Procedure: 1103.018, Maintenance of RCS Water Level

7.0 Development of the Model

No Low Power/Shutdown (LPISD) SPAR model exists for ANal. Therefore, the at-power AN0l SPAR model was modified to allow analysis of the loss of offsite power(LOOP) event. A new event tree (ET) was created to analyze the event.

This ET is shown in Figure A-1 of Appendix A. The ET was linked to a mix of existingat-power fault trees (FT) and new FTs, as applicable. The existing FTs were modified asnecessary to appropriately describe system dependencies during shutdown conditionsand the different success criterion. The ET and high level FTs are shown in Appendix A.

HRA Analysis

Shutdown operation is highly dependent on operator actions as most of the requiredactions are manual (e.g., initiating feed of the RCS). HRA analysis was conducted toproperly characterize the required manual actions. The human error probabilities (HEPs)were calculated using the Low Power Shutdown SPAR-H worksheets from NUREG/CR-6883, "The SPAR-H Human Reliability Analysis Method" and INL./EXT-10-18533 andSPAR-H Step-by-Step" Consideration was given to the available time to perform theaction, the stress levels of the crew during the event, complexity of the diagnoses andactions, crew experience and applicable and relevant training, quality and thoroughnessof procedures, ergonomics, fitness of duty issues, and the available work processes.Table I shows a summary of the dominant HEPs, a detailed discussion of the HEPs isgiven in Appendix B.

In addition to the calculation of specific HEPs for this condition, sequences or cutsetswhich involved multiple operator actions were examined for human action dependency.For the dominant HEPs no dependent couplets were found.

In addition, the cutsets were reviewed to find those that contained two or more HEPs ina single sequence of cutset. For those cutset with multiple HEPs, the HEPs werereviewed to determine if the product of the HEPs was less than I E-6. For those cutsetsa floor, or cutoff, was applied as directed by RASP Manual Volume 4 - ShutdownEvents, Revision 1 Appendix B. A because of the long times to core damage, a cutoff of1 E-7 was applied. This conservative assumption did not materially affect the results

Normal lighting was impacted by the LOOP. This could have an impact on the ability ofthe equipment operators to perform tasks outside of the main control. This impact wasnot assessed.

A detailed description of the HEPs is given in Appendix B.

Page 5

Table ISummary of Dominant HRA Results

S Human Tme Mean. an. TotaError:. Description • I • TDagnmoai AMaon MeanlNeeded Avwilable D IE,,nt ______ _____ sHEP HEP, HEPOperator Fails to Diagnose

SD-XHE-D-LOSDC 5 minutes i 8 hours 2E-5 n/a 2E-5Loss of SDC before boiling i_.

SO-XHE-XL- Operator Fails to RecovereLOSDC Loss of SDC before Boiling ___minutes ____hours na __4E_,_ 4E.4

Operator Fails to Inject (ACSD-XHE-XL-MINJ power available) before Level 30 minutes ' 3 days n/a 2E-5 2E-5

_ Reaches TAF _ _

Operator Fails to Initiate LowSXHE-XL-LPR Pressure Recirc 1 hour 4 days 2E-5 2E-4 2.2E-4

SD.XHE.XM-BWST Operator Fails to RefilWST ST 10 hour 4 days n/a 1 2E-5 2E-5l_ _ _l __l_ _ during Shutdown _ _...... _

8.0 Conditional Core Damage Probability (CCDP) Assessment Results

A detailed Phase 3 Significance Determination Process risk analysis was performedconsistent with NRC Inspection Manual Chapter (IMC) 0609 Appendix G. Step 4.3.8 ofthis procedure directs the analyst to assess the significance of shutdown events bycalculating an instantaneous conditional core damage probability (ICCDP). (Throughoutthis assessment, the analyst has used the terminology of CCDP instead of ICCDP forsimplicity.) This assessment was performed by setting the initiating event frequency(IEF) for loss of offsite power to 1.0 and all other IEF to zero. The above describedSPAR model was evaluated using the SAPHIRE code version 8.0.9.0.

As this SOP evaluates an actual event in which no external events occurred, there wasno risk from external events. As discussed in the above paragraph, this would includesetting any external event IEF to zero also.

The truncation limit was set at IE-!6.

The result of the CCDP analysis is 1.6E-4; based on these results the finding is Red,The top cutsets for are in Appendix C. The analyst did not perform uncertainty analysis.

Table 2CCDP Results

Point Cut SetSequence Estimate Count

4 1.5E-5 83686 1.3E-7 3370a 1.1E-6 25193

11 1.OE-7 63413 .0E-7 13415 1.OE-7 91519 !4E-4 4357

Total 1.6E-4 43171

The results are dominated by two sequences. The largest contributor is from Sequence19 which comprises a failure of the emergency diesel generators (EDG) withoutrecovery. Both the EDG and EDG non-recovery failure probabilities were calculated

Page 6

using the standard SPAR methods and models. Sequence 4 is also a significantcontributor. Sequence 4 cutsets are dominated by combinations of equipment andfailure to recover DHR.

The numeric results above quantify to a Red finding, However, with such a long time tocore damage, recovery is possible with temporary systems such as B.5,b equipment.The analyst is unaware of procedures or training to cool the RCS during theseconditions. In addition, condition in the reactor building will become difficult if not lifethreatening once boiling begins. In conclusion, some credit for these type of actions iswarranted using a SPAR-H approach (note neither SPAR-H nor any other HRA methodwere ever intended to quantify these type of scenarios) would quantify this with a failureprobability between 0.1 and 0.5. If such credit were given, this would reduce the findinginto the Yellow range.

9.0 Conditional Large Early Release Probability (CLERP) Assessment

The figure of merit for this analysis is incremental conditional large early releaseprobability (ICLERP). This ICLERP analysis is based on the method for shutdowndescribed in NUREG/CR-6595 Revision 1, "An Approach for Estimating the Frequenciesof Various Containment Failure Modes and Bypass Events," dated 10/2004. This reportsupplies simplified containment event trees (CET) to determine if the core damagesequence contributes to LERF. NUREG/CR-6595 presents its analysis in terms ofLERF, which is interpreted here as ICLERP.

NUREG/CR-6595 defines LERF as "... the frequency of those accidents leading tosignificant, unmitigated releases from containment in a time frame prior to effectiveevacuation of the close-in population such that there is a potential for early healtheffects," This is identical to the definition of LERF in IMC 0609 Appendix H, Figure 4.2(PWR Large Dry and Sub-atmospheric Containment Event Tree) from NUREG/CR-6595is applicable to the ANO0 event.

This event occurred seven days after shutdown. The earliest core damage could occurwould be three days after event initiation. Thus core damage would not occur until 10days after shutdown. Based on this time and the recommended approach given byNUREG/CF-6595 no large early release could occur.

10. Sensitivity Analysis

Several sensitivity cases were conducted to further understand the event risksignificance. The cases are described below.

Case 1: Loss of Instrument Air

The LOOP event on Unit I in combination with the partial LOOP in Unit 2 combined tocause a loss of instrument air on Unit. There does not appear to be any impact on Unit1 from the loss of air. However, instrument air was being supplied to the steamgenerator nozzle dams. If the nozzle dams had failed, water level could have drained tothe bottom of the steam generator openings. The nozzle dam design appears topreclude a significant inventory on loss of air. The design limits the leakage to a coupleof gallons per minute on each nozzle dam. With several hundred thousand gallons ofwater above the nozzle dams this leakage rate is insignificant,

Page 7

Case 2: HRA No Cutoff

A case was conducted to verify the sensitivity of the results to the cutoff value. Thiscase was run with truncation level of 1E-16. The calculated CCDP was 1,66E-4. Thisindicates that the cutoff implementation is a second order effect only.

Sequence

46811

131519

Total

PointEstimate1.5E-052.5E-081t0E-065.7E-109.5E-134.6E-101,4E-041.6E-04

Cut SetCount83683370

251938341349154357

43171

Case 3: DC Flooding

The stator drop severed a fire water header pipe. It took approximately 45 minutes tostop this leakage. Before the leakage was stopped, water accumulated into the Unit Iand 2 turbine buildings where it caused a small Unat 2 kV fire/explosion. This caused aloss of offsite power to one division of Unit 2 AC power which was mitigated by theassociated emergency diesel generator. Water also started to accumulate into the Unit1 SDC/DHR B pump vault. If this accumulation continued it could have failed the pump.Potentially it could have impacted other Unit I equipment. Sensitivity cases wereconducted with various flooding probabilities and various combinations of impactedequipment. Those combinations and their impacts are presented in the below table.These analyses assume that the flooding could not impaGt the Unit emergency dieselgenerator or their associated 4kV switch gear and 480 v MCCs.

This analysis shows that if the flooding had not been terminated in a timely manner itcould have had a significant impact on plant safety.

CCDP,,,pactd -quip,,, Flood Probability = 0.1 FlooI Probability 1.0'

One LPI/SDC/DHR pump i3E.4 2E-3(either pump A or8) BBoth LPIISDC/DHR pumps 1 E-3 5E-2A single HPI pump (eitherA, B or C) no impact 1.812-4Two HPI pumps (A & B) i" ""no impact . 1.82-4Two HPI pumps (A & C) no impact 2.2E-4

All three HPI pumps no impact 2.2E-4All of HPI and SDCtDHR 1,1E-3 I 2.5E-1

Notes: 1) If the associated basic events are set to True instead of 1.0 the CCDPs aresomewhat lower as would be expected.2) These sensitivity cases were run with truncation set to 1 E-8.

Page 8

Case 4: Impact of Loss of EDG Starting Air Compressors

The LOOP caused a loss of normal EDG starting air. If multiple starts of the EDG wererequired this could impact the restoration of the emergency power. While it is difficult toquantify the change in the EDG non-probability that changes effect on the CCDP iseasily assessed. The analyst assumed that the non-recovery probability was doublefrom 7.14E-2 (for 72 hours) to 1.4E-1. The new CCDP is 2.9E-4. Because the riskresults are dominated by Sequence 19 which is the only sequence effected by the EDGnon-recovery probability, the change in CCDP is directly proportional to the change inthe non-recovery probability on Sequence 19.

11.0 Comparison with Licensee Results

At this time the analyst has seen no licensee results to compare.

Page 9

Appendix A. Model Figures

Page 10

Figure A-I: Loss of Offsite Power Event TreeF ~ ~~~~~ . .....t~lA ... .~Y . W

MOW, tlpp _L-1 Vm cwm vi .r"o ~ 'c ceucJGr~

_ _ _------------ [0 .

0 ~ ~~4 r .......... .

p p IIIIIZZICI:~-F]77OR

Page 1

Figure A-2: Emergency Power Failure Fault TieeI RENERYPOMAER FALS:1E&PS U

FJLU RE: OP~ fidPOWER FROMbWTCW-ZM~ A3

A4CP-$VVGW3 ~ _EAEAUMOIP~ PVR FROM

VACHGM A

ACf-SV4GRt jEd

Figure A-3: Offsite Power Tree

Note that the non-recovery probability was set to one in a change set

Page 12

Figure A4: Diagnose Loss of RHRIDHR Fault Treebm~bg IS LOSSOF RHR

-BEFORE BOUNtiG.

Page 13

Figure A-6: Recovery RHRISDC Fault TreerRocow RJ4RiSOC DURINGSHUTDOWN

SOC-REC

IO*4R 1~A~P~J$JLU~ES

DHR~

II

I

PATH FROM ACS, IET: INL~t MMPLER$IURM4 OHR-

ID o2Z

]PIKWJ3 ?,:, I

I . . .

RCS Suinh oOVCV4 IOAI

DHR T-RAIN.A.FA.S

nVH=R-T11INA

-GHR TMN 13 FIALLS

Page 14

Figure A-6: Gravity and Fo~rced Feed Fault TreeGaft or knoed feed *13ir AM,

pomw tg'LoaifSr)H

t1~NJ -II

I d

WAWI ~FEED.1ATEI3 : 6- 1

Co 0

HFUH PV*ESSURE INJECT1ONduring Shuwdwn

LOW PIESSLtRE WUJECTKQN

Cftwty feed bekyie Coe

G-FEED- EAi

Note the gravity feed portion of this FT is set to fail as gravity feed will not work because the physical level of the BWST is lower than therefueling canal

Page 15

Note this FT is set to fail as gravity feed will not work because the physical level of the BWST is lower than the refueling canal

Page 16

Figure A41: Low Pressure Recirculation Fault Tree

A

6-

Page 17

Figure A-9: BWST Refiil Fault Tree*BVI~rREfla

~FLL

II

R-PL

~'1~I w4w I

0

Figure AIO: Diesel Generator Recovery Fault Tree

ftC14 .t wk.M6RGENCY II DMR-72" I!

HI

Page 18

Figure A-11: SDCIDHR Late Recover Fault TreeLATE ReCOvery b F Si C "ea

:LTRF.C

Note the value of the late recovery basic event varies with the time available

Page 19

Appendix B: HRA Analysis

Page 20

Human Error Probabilities

A high level discussion of the Human Reliability Analysis (HRA) is presented above in Section 7 on ModelDevelopment. Also included above is a summary of the HRA results. The following discusses the HumanFailure Events (HFE), the derivation of the in individual Human Error Probabilities (HEP). This HRA analysiswas done consistent with the guidance of NUREG/CR-6883, 'The SPAR-H Human Reliability AnalysisMethod," dated August 2005.

The Human Error Probabiflties (HEPs) for this analysis were calculated using the Low Power Shutdown SPAR-H worksheets from NUREG/CR-6883. Consideration was given to the available time to perform the action, thestress levels of the crew during the event, complexity of the action, crew experience and applicable andrelevant training, quality and thoroughness of procedures, ergonomics, fitness of duty issues, and the availablework processes.

Page 21

BI Operator Falls to Diagnose Loss of SDC before Boiling

HRA Worksheets for LPSDSPAR HI.PMAN ERROR WORKS lIEIT

Plant: ,NOI Initiating Ewnt: Bmic Ewaf: SD[XIfl-D-l.OSDlWIlasic kWmnt 9scription: Olrator Fails to Diagnose lWs of SD" beforc boiling

Part 1. DIAGNOSIS WORKSHEET

PSaPSIF Levels- Multiplierfur Selected Please note speci11c reasons forDiagnosis PSU rsF lee scection in this

______________________________________ _______ olum",

AMaiJble Tune lnadvcpiutt tirn 5' imnumcs rcequirrd, 8 hours Lavailabiclbrely adoiluae urie (zzl'3 Nominal) r

Exm or 'etv-en I and 2 xtnorninal and -~than 30nin) IL'qnsive (mw (> 2 xcnominal and> 30 mitt

n f rnt inliornmiaion

Conrm1emy HiuhivModcrately. (Zonlek

Ob,, mus diagnesis PUrsM 8toP With l(o55 of power iSns :llcvent infirrrnron ________obvious

ETp'ýr'hrrc Lo~w

Ifigh OI'm uffimecnt inonnalion_______

Procedures Not availbibk

Available, but poor 5Noo-Jnil IXDiagnostcisynmNion1 onentled .

ns~ufficlent imfonmyionExganorrimfl Missing&sMislading .

Poor I')

Insufficient infonyofion1Fitness for Usnfo 1.0& - iDuty Degradcd Rpmes

l.ISUITcient inIronruion

W rk Poorpwucses~ Nwri~nal X

ciod

lnsl-eenitntrntO

Negative PSFs adjustment (_3 negative PSFs)Final liagnosis

HEP

2.00 "S

NA

2100E-5

Page 22

B2 Operator Fails to Recover Loss of SDCIDHR before Boiling

HRA Worksheets for LPSDSPAR4 IRtMAN MRORO WORKS [I=t

Plant. A-NO[ taitimalng Ewatn: Hartic E~tnl: SD-XIIE.XL.IA)SDCBible E~xnf Dt* riptioi Operator Falls to Kecowki lost ofSD{Z before boiling

ftrt HI. ACTION WORKSHEET

Mr~s PSF Levels Multiplier for Selected Please note specific reasonsAction PSF for PSF level selection in

-this column.AvailableTinm Inadequate tim Ti(filure) - 1,0 30minmutes rquid. 8 hours

Tirm Avaeilable is = the owe requiecd 10 avoilabik. SMIXYD IR purrV amcNomir' &~rI cated in the contairimpei one

Time rivaiIble- is Sxth.. tino rotiu rod 0.I N' boiisi occurs into Con~Btam Iut.

Timi avaitahk- A 50.K the linxu requ ffed 0.011 operation of purrips will be emmctdInut itint initbmiation I

St~ress F'Ann 5Hligh 2 X

Insufficient ifrionristion

Moderately 2 x

Insufficrcit irifor-rieion

EWiriiýrnc-e; Tramin i5 Low-M . )~imax

High 0.5Insuiticienit. uIbMattin

Proccdurn; No~t available

Iflcornp i'c 21)Av 11Cbk bar poor 5

Notminal I XInsuffiriet. infonnaiion 1I_____________

Figonomics'1-1,M) Missing/Misk-adinE 50Poor tNnnunlbt I X(.bod 0.5Insufficient inmrrinauttn I

PA;j irDuty Un Futl(~i~r-t

DCgradtd lrisNomaal I X

___________InuiotuIicnt inontu _______IWork Processts Poor

Nornu I~ 1 XGood 0.5

___________________ ~ ~ ~ ~ ~ hya An~fl~n Mirilw __________ ________________

Page 23

B3 Operator Fails to Inject (AC power available) before Level Reaches TAF

HRA Worksheets for LPSDSPAN H1U4MAN IRRORWORH'S[fEF

plant- NMPI blinting Enrt: Baic Ew.nt. SD.XHE.Xl-MINJBasic Ewnt Dwoptitn, Orti rstor Fols to bjec1after UveI Reaches Scram Seloint and before it Reaches TAF

Part 11, ACTON WORKSHEET

PSFS PSF Lev eb Mulipli-erfo6r- Selected Please note ipeclilt reasons forAction PSF PSF level selection in this

_____________ _______________________________column.

Avamdable Time Inadequatc: litec P(faihare)h 1.0Tinv Available is zthe tcmrqvmd1Nuinria: lam~ ITinra available is ý 9x 'he time requir-d 0.1Time available 6- 5 0.x the time nwqorct 0,01 \

ins ufflext~i InfortraLkon

High2 X

Ins uff~iciril ifnnatioitknICurnlikoxty Highly 5 'ihiý s ciomes that condens ate

Moderately 2Conlainus to run on lWS oTVC. 1I"111lII x ticking .m core spray is requited Lhi3

Insufficicnt1 inftrniairt I - Wvrtld be rTodTtwc.

F-Vfirmene,!lranine I.Dw 3

HighVIns ufficient infonnaicnI

Pro,.cdurts Not availlableNIntcarirplole 20A vallabi but poor5

NimitntelI xInsufficicrnt infomnatio _______I ____________

Es~no ~ Misl shingi/Mnsleading 5Poor

InsuflieneIs inforia ton ______I -

Fintess for Duly Unfk P(I'aflurc)~ 1 .0Dcgrndcd Fiv5

Irsuffcieitt ionfmitonIWork 11roeysses Poor

Good(1Inxtificient ninibrithon10r

Negative PSFs adjuastmnen; LaS negative PISFs)

Minal Action If £P

INA

Page 24

B4a Operator Fails to Diagnose Need for Low Pressure Recire

HRA Worksheets for LPSDS PAN HUMAN EHROR WORKS IfHWf

Plant: ANOI Initiating Ewnt: Basic -vent; SDA-XMML-LPRBask le.nt Descripton: Operator Fails to Initiate LowlPressure Rtcirt

Part 1. DIAGNOSIS WORKSHEETMr~s PSF Levels Niultiphier 11r Selected Please note specific reasona br

Diagnosis PSF PSF level selection in this________________________________________________colum~n.

7vailable fine Inadequate Linz PQt .0c =t Feed has been startd Naen to thereBarely adequatc tirm (:t2.'3 Normnal) 1(3 is at least 24 hours to rattanx SlXCNormiwi iwcFxra tinic (betwvun I and 2 x norninal at-.d > than 30 inin) 0.1Exprasie Ike (> 2 x horinal and > 30 nrnn) 001

_______Insuflieret Inonio ______

strias Exrrne 51 figh 2X

Insuffleicrit inlonrationIConvikxgy lijghly

Mokrately, conv le 2Nominal I x

Obvious diagnosis0In~allickrit inibmnertton I srarn setwornt is an obvious cue

LnP~entCt, to7 10Training .wn. nal Ix

Insufftkrint infornit1ionIPt'occdurecs 01t aveilablC so

ric:nijiatc20Avijlable, but poor 5

Nominnall xOiagnostiusysvn'ptomnorsenwd 0.s

Fr69"ornics/Il is s ingM is leuadin& 3Peor 10

Insufficient infurniton IFitness 1b, Untlt P~fadurg) - 10Dowt De~raded Fitnss

linsufficieni inforoiwiionWorki Poor2

JI.,ufkicont i~n farr200 Io

NUEP - 2-OOE-4

Negative PSFs adjustment (!3 negative PM~s) NAFinal Dhiag rums HEP = 2.OOE.4

Page 25

B4b Operator Fails Action for Low Pressure Recirc

HRA Worksheets for LPSDSPAR HIIMAN ERKOR WORK-SHE;r

Plant: ANOI Initiating E•nt: Basik BmattSD-IlE-XL-UlAiBamic &vnt Ncription. Operator Fails to Iiniat Low Pressure Kecire

Pantif. ACTION-WORKSHEET

PSFs PSF Levels M Uttiplic r fokr SelCcfted Please note specific reasons forAction PSF VSF level selection in this

Avaiiable Twn Inad~uquc firre Wafiurc1" 1t)Tinw Avatlahk is zthe tint~ rcqu;rcd 10Nomnirini zmrýT inx available is -e S5the dift required 0.1Tim avai~lable is; 50Sx the irrma required 0.01Insufloc~ientm at

High 2

JnSU11fiejnt ~frirGrjmie*k Hmghky S

Modcramedy2

I 1sufricsoni t mtmmF.1wnicnwm~ning, Low

High 0.5Insufficmcnmt mforn~iiotnI

Pro.-Aures Nont mmvailablt 51Incfl tIeI 20Amiabl~e but poor 5.4rm~ii 1 XInsufficmen infomzask',nI

ln~ginnrmitsHMl Mi~sg!Mtsleading 50PowoI

MiimlI 'mx

insuffimcint intim~am~nh________Fitness for Duty UnNi P 1.~e =I0

Dcgrmdod FitncisNvninad I X

_________Insuffrmenic mafrtiion I_____

Work Proce~ics poor

Good 0.5]Isu fficientmfi mt aiomIwn 11

• .. ':l•: :. , i ,. " ." , . .i,..,

NHEPNegative trSrs adijmtment (>3 negative FPSl~)

F~inal A crin HE5P

2.00 ~.4I5

NA

2(10 E.OS

Page 26

Appendix C: Cutsets

Page 27

Top 20 Cutsets:

Page 20

Fairbanks, Abin

From:Sent:To:Subject:Attachments:

Tindell, BrianTvesday. November 19, 2013 2:21 PM

Young, Matt; Fairbanks, AbinFW: prelimianry draft of ANO Stator Drop SDP Draft Revison 0ANO LOOP SDP Analysis Rev O.0.docx

FYI - I read through this. Preliminary Yellow just for the stator drop.

From: Werner, GregSent: Tuesday, November 19, 2013 3:03 PMTo: Bloodgood, Michael; Melfi, Jim; Tindell, BrianSubject: FW: prelimlanry draft of ANO Stator Drop SDP Draft Revison 0

FYI

From: Weeralkody, SunilSent: Tuesday, November 19, 2013 11:46 AMTo: Loveless, David; Miller, Geoffry; Wemer, GregCc: Mitman, JeffreySubject: prellmianry draft of ANO Stator Drop SDP Draft Revison 0

From: Mitman, JeffreySent: Thursday, November 14, 2013 4:32 PMTo: Weerakkody, SunilSubject; ANO Stator Drop SDP Droft Revison 0

U-' ~ v\O~/\~c l\kAf\r.

Sunil, attached is the subject analysis and a zip file containing the SPAR model for review and comment. Thezip file may be too big to email to Region IV. I'm still working on the SDP on the loss of SDC at midloop.

Jeff Mitman

1

- Torbxy, Andrea..... ...... ... . _ . .. i

From, esn os r 11 7e- e A11 '4

Sent: Fr~day, April 05, 2013 5:49 PM .5 S, e ,,

To: Kobetz, TimothyCc: Lewin, Aron; Cauffman, Christopher, Klet. Audrey, Levasseur, Gabiiei; IsoaT, James;

Campbell. Stephen; Cartwright William; Gamberont, Marsha

Subj.1 FW: ANO MD 8.3Attciment5 MD 8 3 for ANO stator drop Rev3.docx

FYI - ANO AIT... ...... ...... ... ... . .... ... ..... ........... . ." .... ... . .... ... ..... ..........; ........ .... ....... .. ... ... ....... .. ... .. .... .. ............. .... i ... .... .. .. ........ ... ............ :. i ............. .... i ........ ... ...... ... ...; ............... ...T ... .. : .... .. ...... .... ...... ................ .............. ...: ....... .... ... ... . ... ... ................

From, Pannier, Stephen f/4,,. 1w r , ', .. , '.,.r k, /o ;•,.,. / , ,Sent, Friday, April OSi 2013 9:36 AMTo: Nieh, Ho; fH:e, AllenCc: King, Mark; Markley, Michael; Telson, Ross; Sigmon, RebeccaSubject: FW: ANO MD 8.3

Attached is a copy of the AND MD 8.3 deterrnination,

Thanks

Steve

from: Allen, DonSent: Friday, Api' 105, 2013 9:25 AMTo: Wang, Alan; Chernoff, Harold;. Weerakkody, Sunil; Balazik, Michael; Jones, Steve; Mendiola, Anthony;. Panniet,Stephen, Loveless, David; Clark, Jeff, Blount, Barbara; Garmno, DavidSubject. FWt ANO MD 8.3

For the diScussion at 9:00 cenlral time today

. . .... . .. . .. . . . . . . . . . .................. . ......................... . ..... . . ... .. .. ...... . .. .. ... . . . .. . . .. ......... . . .... ......... . . ... . .... . .. .. ..... ............ . . ...... .. . ..... . .... . ....... .. .. ........... ... .. .... ... ... ........ ............. .... . . .. . . , . . .. • • . , . . : . .. : . . .

From: Lusk, Rustin~..Sert. Friday, April 05, 2013 8.-18 AM

To:. Maike,: MichaelCc: Allen, Don; Miller, GeoffreySuibject: ANO MD 8.3

Good Morning,

Please see the attached report. Thank you.

Respectfully,

Rustin "Russ" LuskDivision of Reactor Projects, Drviston Admiin AssistantU.S. Nuclear Regulatory Commission, Region IVPhone: (817)200-1184Fax: (81.7) 200-1278

760.• is baing sure of wbAhl we do not see & .ct.in of whal we hope for."R.LP. Rick 'RpperRy,•enl*

2

UNITED STATES•" il•-,:d •NUC LEAP. REGULATORY C OMMI S StON

'* ... REGION IV

INO EAST LAMAR SLVOARLINGTON, TEXAS 76W11.4511

April 5, 2013

MEMORANDUM TO: Arthur T. Howell 111, Regional Administrator

THRU: Kriss M. Kennedy, Director, Division of Reactor Projects

FROM: Donald B. Alien, Chief, Reactor Projects Branch E

SUBJECT: MANAGEMENT DIRECTIVE 8.3 EVALUATION FOR ARKANSASNUCLEAR ONE

Pursuant to Regional Office.Policy Guide 0801, "Documenting Management Directive 8,3Reactive Team Inspection .DecisiOns," the endosed table provides the Management Directive0.3 evaluation of the March 31, 2013, event at Arkansas Nuclear One involving the failure of theUnit 1 main generator stator lifting rig. Based on the results of the MD 8.3 evaluation (attached),I recommend that we conduct a:n augmented inspection at Arkansas Nuclear One.

Concur with Recommendation:

Arthur T. Howell Ill, Regional Administrator

Enclosure:MD 8.3 Decision Documentation Form

L A. Howell III

Electronic distribution by RIV:Regional Administlator (Ant.Howelt@nrr,.gov)Deputy Regional Administrator (Robert.Lewis@nrcgov)DAP Director ([email protected])Acting DRP Deputy Director (Michael. [email protected])DR.$ Director ([email protected])Acting DRS Deputy Diector (Jeff.Ctark@nmcgov)Senior Resident Inspector ([email protected])Resident Inspector (William. [email protected])Branch Chief, DRPIE ([email protected])Senior Project Engineer, DRP/E ([email protected])Project Engineer, DRPIE ([email protected])Project Engineer, ORPWE (Dan.Bradley@nrc. gov)ANQ Administrative Assistant (Gloria.Haztfietd@nrc. gov)Public Affairs Officer (Victor. Ddcks@nrc, gov)Public Affairs Officer ([email protected])Prjpecat Manager ([email protected])Branch Chief, NRR/DRAIAPOB ([email protected])Branch Chief, NRRIDlFSIiOEB (Harold. Chernoff@n rc.gov)Branch Chief, NRPJDORL/LPL4 (Michael.Markley nrc.gov)Branch Chief, DRSTSB (Ray. Kellar@nrc. gov)ACES (R4Enforcement. [email protected])RITS Coordinator ([email protected])Regional Counsel (Karla.FulIer@nrcgov)Technical Support Assistant (Loretta.Wi lliams@nrc .gov)Congressional Affairs Officer ([email protected])DRP Director, Region I ([email protected])DRP. Director, Region I1 ([email protected])DRP. Director, Region III ([email protected])

R1, MD &.3 Decisions, ADAMS MLSUN.S Rev _.mp.. lA.Yes 0 No 'ADAMS .Yes 0 No o Reviewer Initials OBA

Publicly Avail Yes No Sensitive Ye 0 o Ses. Type Initials OBAPublic AVail Date Keyword MD 3.41A.7

t

ii

DAllen TIBlount K~nd ________

__________________7 1 ~I~

.OFFICIAL RECORDJ COPY T=Telephone E--E..rai F=Fax

MD 8.3ttMC 0309DECISION DOCUMENTATION FORM

L (Deterministic and Risk Criteria Analyzed)rr.. . . . . .. .1.. . . . . . ..

PLANT EVENT DATE 3/31/2013

RESPONSI1LE EVALUATION 4/1(2013MRANCH CHIEF Don Alen DATE

BRIEF DESCRIPTION OF THE SIGNIFICANT OPERATIONAL EVENT OR DEGRADEDCONDITIONn March ý31, 2013. at approximately 7:50 a~m. (CDT), Arkansas Nuclear One (AN%) Unit 1 which was

in a refueling outage, lost Doffste power and ANO Unit 2 experienced a reactor trip after a-600 toneneator stator ftl onto the turbine deck and then approximately 30 feet onto the train bay: floor. The

eectrical non-vitai busses supplying offsite power to Unit 1 were damaged, and some of thefiresuppression system piping was damaged. The falling stator and cyane components caused the svppiy.breakerto:Unit Z reactor coolant pump 8, o open. The loss of reactos coolant pump B resulted in a Unit2 reactor trip, whichi had. been operating at 100 percent power. Both units are stable and remainshuidown.

....The licensee reported that one worker was killed and ebIt others were injured wlien the main:generatorOf . stator fell. Seven workers have been treated and 'eleased from a hospital, vAAse one remains°..•. ! hospitalized.

With the loss of offstte power to Unit 1, both Unit 1 emergency diesel generators (EDGs) started andloaded onio their electrical busses, Decay heat removal was quicklyresored. The Unt 1 emergency

jeset generators continue to supply power to the vital electrical busses.

Al. 9-22 am., (CDT), offsite. power to Unit 2 from startup. transformer 3 was lost because water from a firemain caused a short cbrcui.. ANO Unit 2 EDG 2 started and energized the train B vital electr,'da bus,while traln:A vital and non-.vital electrical busses wefe re~energizod from startup transformer 2. Thesupply tbreaker from steitup transformer 3 failed because -of water intrusion sterMming: from damaged. fireuppression system piping. Operators cooled down Unit 2 to hot shutdown.

At 10:44 a.m. (CDT), the licensee declared a Notification of Unusual Event because the failure of theLupply breaker may have been caused by an explosion in the breaker cubic4e. The event was

terminated at 6,21 p.m. (CDT) because the affected electrical bus was not energized and there was nother damage. The fire suppression system to ANO Unit is shutdown due to the damage to the fire

Water. system. piping. Damaged portions of the ANO Unit 2 fire protection sy'sem have been isotated:Additional fire water pumps have been positioned to provide fire water if necessary. The licensee hasestablished fire watches in the auxiliary buildings of both units.

The deterministic criteria described in MD 8.3 and MC 0309 was reviewed, and the criteria fisted belowere determined to be applicable to this event.

-1- Enclosute

-V

Y/N DETERMINISTIC CRITERIAd. Led to the loss of a safety function or multiple failures In syst.s used tomitigate an -actual event

Remas- The failure of the lift system resulted in the main generator stator damagingthe electrical buses supplying :offsite power to Unit 1. The damage

y resulted in Unit I losing both trains of offsite power, Both EDGs fdr Unit 1started automatically and are supplying power to their buses.. The Unit 1fire suppression system was damaged :and during the event and portionsof the system are secured. A portion of Unit 2's fire water sys!tem wasdamaged and caused the feeder breaker from. a startup transformer to_open, This resulted in a partial-loss of offnite power to Unit 2.

e. Involved possible adverse generic implications

y - Nuclear power plants conrduct lifts of Naevy equipment from tWme to .time.Although unknown at this time, the cause(s) of the failure of. the lifting rig:could have adverse generic implications,

-2-

15CO DITANALY RI aidLvlSsKAE ApSSESSME201IF IT IS DETERMINED THAT A RISK ANALYSISIS NOT REQUIRED- ENTER NABELOW AND :CONTINUE TO THE DECiSiON BASIS BLOCXTK

RIKANALYSIS BY- David Loveless IDATE- April 1, 2013

Brief description for the basis of the assessment:

The senior reactor analyst evaluated the risk :of this event using the ArkansasNuclear One, Unit 2, Standardized Plant Analysis Risk (SPAR) model, Revision8.21, Inspection Manual Chapter 0609, Appendix G, Attachment 2, and otherqualitative: assessment tools.

The analyst assumed that the event in Unit 2 was similar to an uncomplicatedreactor transient with Switchgear 2A2 out of service. The resulting conditional coredamage probability (CCDP), 1.1 x 10 -6, indicated the lower bound of the risk fromthe drop. Assuming that the risk could be bounded on the high side by modelingthe event as a plant-centered loss of offsite power. the CCDP was quantified ast.3 × 10-5.

The analyst used Figure 8 from Appendix G, Attachment 2, to assess the risk of thedrop event on Unit 1. The ficensee informed the analyst that one of the breakersrequired to power the vital busses from the aWternate.ac diesel generator was notavailable beoause of potential damage from the event, Therefor. the analystcalculated the probability of an emergency power supply system demand failure at4.49 x 10-3, assuming that only Diesel Generators I and 2 were available to supply:vital loads. Given that offsfte power had not been restored within 36 hours and was.not expected to be returned for some time. the analyst set the probability of :failureto restore offsile power to 1.0, The probabi!ity of not recovering a postulated dieselgenerator failure within 18 hours was derived using the SPAR as 3.63 x 1:0-1, Theanalyst used a screening. value .of 0. 1 for the probability of alternative strategiesfailure leading to core damage. The resulting CCDP was 1.6 x 10-4.

The anialyst noted that thefe were several unknown aspects of the event that couldaffect the risk. These issues are listed below:

Unit 1:

The failure probabil.ity for the emergency diesel generators was set using aprobability for failure-to-run for 24 hours. As the diesels are demanded for longerperiods of time, the probability of failure to run increases faster than the probabilityof recovery. Therefore, this probability would suggest an increased CCD:P.* Configuration and operation of the instrument air system would have animpacton the probability of a loss of fnventoy from the refueling pool.

-3-

Il CONDITIONAL RISK ASSESSMENT .Condition Report indicales that there was a loss of all instrumentation

related to the nozzle dam seals.The ability of operators to vent the containment following boiling in the

refueling pool would have an impact on the success of alternative strategies.

Unit 2:

The condition and structural integrity of the walls and ceiling of the TurbineBuilding 362 - foot elevation switchgear area can affect the probability of continuedoffsite power to the unit

Peer Review:

The analyst's results were peer reviewed and concurred upon by analysts.from NRRJDRAJAPOB.

Licensee:

The analyst discussed this analysis with 1he. licensee's anaiysls, The

licensee does not have a shutdown risk model. Their qualitative risk tooi indicatesthat a loss of offsite power while shutdown is a Red condition. During the in-itialdiscussion,. the licensee stated that the evaluation seemed reasonable for Unit 1isk., but requested that we give them mare .credit for the available time to recover

:power.

The licensee provided a conditional core damage probability for the Unit 2event of 2 x 10-7. The licensee did not consider Bus 2A2 to be unavailable, nor didthey account for the condition it was in at the start of the event. Additionally.. thelicensee did not account fof the secondary equipment that was not poweredbecause of Diesel 2-2 supplying Bus 2A4.

Note.: description may include assumptions, calculations:, references, peer review, or comparison withi!ernsee reeJtts.

THE CONDITIONAL CORE DAMAGE PROBABILITY (CCDP) 1I Unit 1: 1..6E.4Unit 2: 1.1E-6 1.36-5. ..iCH JPLACE. JJ. . ........T --

WHICH PLACES THE: RMS IN THE RANGE OF

-4-

RESPONSE DECISION AND-BASISUSING THE ABOVE INFORMATION AND OTHER KEY ELEMENTS OF

CONSIDERATION AS APPROPRIATE, DOCUMENT THE RESPONSE DECISION TOTHE EVENT OR CONDITION, AND THE BASIS FOR THAT DECISION

.. . . . . . . . . . . .... . . . . . .

Response to the.avant or Augmented Inspectioncondition

BASIS FOR. THE RESPONSE

Ba•ed. on meetng the two deterministic criteria. and the results of the conditional, riskassessment (including the uncertainties associated with assessment), I have concluded.thiat the NRC should conduct an augmented inspection at Arkansas Nuclear One.Information gathered during the inspection will be evaluated to determine if an augmentednspection, Is the appropriate response to this event.

COMPLETED BY )onaid B, Allen DATE

BRANCH CHIEF )onald B, Allen DATEREVIEW:

DIVISION <riss M. Kennedy DATEDIRECTORAPPROVAL

A

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